ML20205D490
| ML20205D490 | |
| Person / Time | |
|---|---|
| Site: | Perry, River Bend  |
| Issue date: | 02/22/1999 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20205D487 | List: |
| References | |
| 2.206, NUDOCS 9904020221 | |
| Download: ML20205D490 (188) | |
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F 1 1 ' UNITED STATES OF AMERICA 7[) 2 NUCLEAR REGULATORY COMMISSION q 3 *** 4 ' PETITIONS' PURSUANT TO 10 CFR 2.206
'5 PERRY NUCLEAR POWER PLANT AND RIVER BEND STATION
'6 ***
7 INFORMAL PUBLIC HEARING 8 *** 9= Nuclear k?gulatory Commission 10 Room T2-B3 11 Two White Flint North
- 12 11545 Rockville Pike 13 Rockville, Maryland 14 15 Monday, February 22, 1999 16 17 The Commission met in open session, pursuant.to I 18 notice, at 12:59 p.m., ELINOR ADENSAM, Perry Nuclear Plant
-19 Project Director, presiding.
20' Remote Teleconference Sites: Louisiana State 21 University in Baton Rouge, Louisiana, Lake Erie College in 22 Perry, Ohio, and a site in Cleveland, Ohio. 23 24
.25 ANN RILEY &' ASSOCIATES, LTD.
Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
l^ 2 1 STAFFcAND PRESENTERS:
]'~$. 2
(_/
' i -33 ELINOR'ADENSAM,TProject Director, Project Directorate'I-2,
~4 . Office.of Nuclear Reactor Regulation 5 . DAVID.LOCHBAUM, Union of Concerned Scientists +
~6; JERRYrWERMIEL, Branch Chief, Reactor Systems Branch I
- I 7- JOSEPH DONOGHUE, NRC Reactor Systems Branch j 8 RALPH CARUSO, NRC Reactor Systems Branch .
)
9 . JOHN HANNON, NRC Division of' Licensing & Management 10 STU-RICHARDS,'NRC Division of Licensing & Management i 11 CHAFLIE RICHARD, Branch Chief, Region IV
'12 GEORGE REPOGEE, Senior Resident Inspector,-River Bend 13 JEFF CLARK, Resident-Inspector, Perry i
14 CHRISTINE'LIPA, Senior Resident Inspector, Perry i m
.Q 15 : STEPHEN BETHAY, Entergy' Operations, Incorporated 16 RICHARD KING, Entergy Operations, Incorporated I
17 FRED TITUS, Entergy Operations, Incorporated ) 18- LEW MYERS, FirstEnergy Nuclear Operating Company 19 -HOWARD BERGENDAHL, FirstEnergy Nuclear Operating Company 20 HENRY HEGRAT, FirstEnergy Nuclear Operating Company 211 PAUL BORDLEY, FirstEnergy Nuclear Operating Company l 22 RUSS KEARNEY, FirstEnergy Nuclear Operating Company 23 PARTICIPANTS (Continued):
'24 2 5 ._ MARY O'REILLY, Esquire, Corporate Counsel, FirstEnergy
.O ANN RILEY & ASSOCIATES, LTD.
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i I 3 1 J 1 Nuclear Operating Company [ 'i '2 JIM EMLEY, FirstEnergy Operating Company _
- u-). j 3 ROBERT FRETZ,LRiver Bend Project Manager l 1
4- DOUGLAS PICKETT, Perry-Project Manager
-. 5 6 1 7-8 9- --
l 10 q 11 12 l 13 14-() .es 15 16-17 18 19 20 l 21- - 22 j 23 24 25'
)
ANN RILEY & ASSOCIATES, LTD. I ("'/I s- Court Reporters {
'1025 Connecticut Avenue, NW, Suite 1014 Washington,,D.C. 20036 .
(202) 842-0034. l
r l 4 ) 1 1 PROCEEDINGS l q
; ; 2 [12:59 p.m.)
N._) 3 MS. ADENSAM: Good afternoon. My name is Elinor 4 Adensam. I am the Project Director of Project Directorate 5 I-2 in the Office of Nuclear Reactor Regulation. I have 6 been designated by the Director of that office to chair this 7 meeting, called an informal public hearing. 8 We are meeting this afternoon to obtain additional 9 information related to two petitions submitted by tir. David 10 Lochbaum of the Union of Concerned Scientists pursuant to 10 11 CFR 2.206. The two petitions concern the operation of the 12 River Bend Station, located in St. Francisvil?.e , Louisiana 13 and operated by Entergy Operations, Incorporated, and the 14 Perry Nuclear Power Plant in Perry, Ohio operated by l
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t s_ 15 FirstEnergy Nuclear Operating Company. i 16 The sole purpose of this hearing is to gather
)
l 17 information related to these petitions. The petitioner, the 18 licensees, and the public will be afforded an opportunity to 19 speak. The information provided today will be considered by 20 the NRC Staff in its evaluation of those petitions. Because 21 this is an informal hearing intended solely to gather 22 information, the NRC Staff will not necessarily respond to l i 23 comments or questions from either the petitioner, the i 24 licensees or the public. This lack of response should not 25 be inferred to indicate either agreement or disagreement [ ANN RILEY & ASSOCIATES, LTD.
\- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 '
5 1 with statements made by any of the parties. We are here (,~,\ 2 merely to listen and gather facts, nor are any of the b) 3 parties expected to respond to remarks or questions from 4 anyone outside of the NRC Staff. 5 However, if a party believes an issue can be 6 cleared up simply by a factual response and they wish to 7 make sure a response, they may volunteer to do so. 8 Questions by participants which are not answered 9 here will be responded to by the Staff in the appropriate 10 forum to the best of our ability. 11 This hearing is being videoconferenced at three 12 remote locations for he purpose of providing the 13 opportunity for wider public participation. Those sites are 14 Louisiana State University in Baton Rouge, Louisiana, Lake I ,~\ (s-) 15 Erie College in Perry, Ohio, and a site in Cleveland, Ohio, i J 16 While this is not the first of these informal hearings that 17 the NRC has conducted, this is the first one where we have 18 attempted to videoconference with remote sites. As a result j i 19 the logistics are more challenging and will require l 20 interested parties to cooperate, especially with respect to 21 time constraints. 22 We want to permit persons who want to to have the 23 opportunity to speak. We also want to gather as much 24 information as we can related to the petitions to assist us 25 in our evaluation. Unfortunately, we will lose the remote [~ \~ '
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m 6 i 1 location in.-Baton Rouge.at'4:45 p.m. Eastern Time and the- ,, .2 location in-Perry, Ohio at 5 10 p.m. Eastern Time. 3 Therefore, before I-go into introductions I would like to . 4 review conduct of this informal hearing. q 5 Following'these introductory remarks and the 6 introductions, representatives from the Union of Concerned i a 7 ' Scientists will be allotted'approximately 45 minutes to 8 provide additional information related to the basis for 9 their. petitions. The NRC Staff will then be allotted 10 approximately 15 minutes to ask questions of the petitioner 11 for the purposes of clarification. Each licensee will then l 12 be allotted approximately 30 minutes to address issues 13 raised in the petitions. The NRC will be allowed 14 approximately 15 minutes after each licensee's presentation ( 15 to ask questions for the purpose of clarification. 16 Following this, we have allotted approximately 45 minutes l 17 for the public to provide comments relative to the petition. 18 If additional time is necessary, we may adjust the ] 19 schedule to allow for public comments. As mentioned' 20 earlier, due to the classroom scheduling needs, participants 21 at LSU and Lake Erie College will need to leave their 22 classrooms at 4:45 Eastern and 5:30 Eastern, respectively, 23 thus the order for public comments will be, first, 24 participants at LSU, second, participants at Lake Erie, 25 third, participants in Cleveland, and then those ANN RILEY & ASSOCIATES, LTD. O- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
7 q 1 participants here in Rockville. Should there be more [') v ' 2 participants than we have time for at the remote locations, 3 especially those that we lose early, we ask that they speak l 4 to the NRC representative at those locations and provide 5 their comments to the Staff member so that we can consider i I 6 them in our evaluation of the petitions. 7 In addition, they will be included in the meeting 8 summary but obviously not in the transcript. l l 9 Finally, approximately five minutes each will be 10 allowed for the petitioner and the individual licensees to 11 make closing statements. 12 This hearing is being transcribed to produce a 13 formal record. That record will be made available to the 14 public. Persons attending the hearing here in Rockville and f (3,/ 15 those at the remote locations should sign the attendance 16' sheets and indicate on the sign-up sheet whether or not they 17 will want a copy of the transcript. 18 Because this hearing is being transcribed, it is 19 very important that persons spec:.ing state their name 20- clearly so the Recorder can record it correctly. If the i 21 spelling is unusual you may also want to spell it. 22 Names of the persons at the table have been 23 provided to the Recorder in advance, but it would not be 24 remiss to introduce yourself the first time you speak. 25 There are microphones available here in Headquarters and at / ANN RILEY & ASSOCIATES, LTD. '%s Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
8 1 the remote sites. Please use them to aid the Recorder in 2 hearing what is said. (V) 3 I would like to go through introductions and I i 4 would like to introduce,.first, Mr. David Lochbaum of the { l 5 Union.of Concerned Scientists, who will be presenting 6 addition'1 information related to the two petitions that l 1 7 have been filed. l 8 Mr. Lochbaum, you said you have to staff members 9 with you? 10 MR. LOCHBAUM: I am by myself today. 11 MS. ADENSAM: Thank you. I would like to 12 introduce Mr. Titus, Fred Titus, who is Vice President of 13 Engineering. He'll be speaking on behalf of Entergy 14 Operations, Incorporated. g ) 15 Mr. Titus, would you introduce your staff, please? 16 MR. TITUS: Yes. I am Fred Titus, Vice President 17 of Engineering for Entergy Operations, the operator of the 18 River Bend Station. 19 To my direct right is Mr. Rick King, Director of 20 Nuclear Safety and Regulatory Affairs at the River Bend 21 Station. Further to the right is Mr. Steve Bethay, Director 22 of Licensing for Energy Operations, and in the back we also 23 have some support personnel that for the purposes of time I 24 will not introduce the individuals in the backup row. 25 MS. ADENSAM: Thank you. Mr. Lew Myers, Vice 1
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C Court Reporters , 1025 Connecticut Avenue, NW, Suite 1014 l Washington, D.C. 20036 (202) 842-0034 j
f I J 9 1 President of Perry Nuclear Plant -- he is going to be (,~) 2 representing FirstEnergy. Mr. Myers, would you introduce
%./
3 your staff, please? 4 MR. MYERS: Yes. Thank you very much for the l 5 opportunity to be here today. l 6 I would like to introduce my staff. I am Lew 7 Myers. I am the Site Vice President of Perry Nuclear Power 8 Plant. To my left is Howard Bergendahl, who will help lead 9 some of the comments -- the discussion today. Howard is our 10 Director of Nuclear Services. 11 To his immediate left is Henry Hegrat. Henry has 12 an SRO in our plant and 23 years of experience. He is with 13 us today and he is in charge of our licensing group. 14 Behind me I have Paul Bordley, a graduate of MIT, I () fm 15 operations superintendent, is with us today and Mary 16 O'Reilly, our corporate counsel, is in the back, and Jim 17 Enley, one of our senior licensing engineers came to us from 18 the United States Navy and has been with us for 23 years. 19 MS. ADENSAM: Okay, thank you. I am going to ask 20 the NRC Staff here in Headquarters to introduce themselves, 21 and then I will introduce the NRC representatives at the 22 remote locations and so they can be seen by people here, 23 when I introduce you if you would please respond verbally so 24 that your picture will show up on the video for us, we would 25 appreciate it. (~ ANN RILEY & ASSOCIATES, LTD. (_s Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
T 10 1 I'would like.to start with you, Mr. Donoghue. [' ?) 2 MR. DONOGHUE: 'I am Joe Donoghue. I am in Reactor A.J . 3 Systems. I am involved in the-technical review. 4 'MR.TCARUSO: My name is Ralph ~Caruso. I am in the 5 Reactor SystemsfBranch involved in this review. 6 MR. HANNON: I am John Hannon, c'he Project
-7 Director in NRR responsible for licensing at River Bend.
8 NR. RICHARDS: I am Stu-Richards. I am'a Project'. 9 Director at NRR and'I am responsible _for licensing at the 10 Perry plant. 11 MS. ADENSAM: Mr.'Wermiel? MR. WERMIEL: My name is Jerry Wermiel. I am the 13 new Chief of the Reactor Systems Branch in the Office of 14 Nuclear Reactor Regulation. () 15 MR. EICKETT: I am Doug Pickett. I am the NRR ) 16 . Project Manager for ths Poiry plant.
=17 MR. FRETZ: I am Robert Fretz. I am the Project 18 Manager for River Bend Station.
19 MS. ADENSAM: Thank you. At the LSU location, we 20 have Mr. Charlie Marshall from our Region IV office. 21 Charlie, could you speak? 22^ MR. MARSHALL: Hi. I am Charlie Marshall -- 23 MS. ADENSAM: And also is Mr. George Repogee 24 .there, please? Is he with you, Charlie? 25 5 MR. REPOGEE: Hi, This is George. I'm the senior n ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
l 11 L 1 resident inspector.at River Bend. I - 2. MS. ADENSAM: Okay,'thank you. At Lake Erie I ( ! ~31 . College,-~we have Mr. Jeff Clark, who is the resident 4: inspector at Perry. Jeff, could you speak? E '5 [ Pause.]'
~6 MS. ADENSAM: Jeff, are you not miked? Do you.
7 have a mike? Because we are having difficulty hearing.you. 8 [ Pause.] 9 MS. ADENSAM: With difficulty. 10 [ Pause.] 11 , MS. ADENSAM: Better, yes, thank you. Can you 12 hear? No. I am sorry, Jeff. Our Recorder cannott hear you.
- 13. MR. CLARK: Can you hear me?
14 MS. ADENSAM: You can't hear it? Okay. Yes, but
.g-() 15 if you do have people speaking there, they are going to~have 16 to speak up. Thank you.
1"i . In Cleveland we have Ms. Christine Lipa, tr-18 senior resident' inspector at Perry. Ms. Lipa. 19 MS. LIPA: Can you hear me? 20 MS. ADENSAM: With difficulty. I 2 11 MS. LIPA:- Is this better? This is Christine 22 Lipa:-- i 23 MS. ADENSAM: Yes, yes. Thank you. 24 I would also like to acknowledge members of the 25 public that we.have attending at each of these locations as 1 l l l l [' A ANN RILEY & ASSOCIATES, LTD. Court Reporters I
.4, 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 l 1
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12 1 well as those people here in Rockville. We welcome you. We n (% 2 look forward to any information you have to shc
,- ) with vs 3 relative to these petitions.
4 The Agency has provided a mechanism in 10 CFR 5 2.206 whereby intere:sted citizens may petition for agency 6 action related to the facilities that it regulates. 7 The Staff guidance for reviewing such petitions is 8 provided in the Agency Management Directive 8.11. That 9 directive provides for holding an informal public hearing to l l 10 obtain additional information from the petitioner, the
]
l 11 licensee, and the public for NRC Staff use in evaluating the 12 petition. It is not a forum for the Staff to offer any 13 preliminary decisions on the evaluation of the petition. 14 The two petitions submitted by Mr. David Lochbaum I g () 15 acting on behalf of the Union of Concerned Scientists 16 pursuant to 10 CFR 2,206 inde$endently requested that the l 17 River Bend Station and the Perry Nuclear Power Plant be I 18 immediately shut down and their operating licenses suspended 19 or modified until each facility's design and licensing bases 20 were updated to permit operation with failed fuel assemblies 21 or until all failed fuel assemblies were removed from the 22 reactor core. 23 In addition, each petition requested an informal 24 public hearing in the Washington, D.C. area to presant new 25 plant-specific information regarding the operation of the
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[V Court Reporters 1025 Connecticu[ Avenue, NW, Suite 1014 Washington, D C. 20036 (202) 842-0034 i.
L 13 i L 1 named facility as well as to discuss an April 2nd, 1998 UCS I [ ] 2 report entitled, " Potential Nuclear Safety Hazard Reactor
\ss' j
3 Operation with Failed Fuel Cladding." j 1 4 Although in each case our preliminary views, based 5 on our knowledge at the time and concluding that no 6 immediate action was warranted, were transmitted to Mr.
*/ Lochbaum, those views do not prejudice the finai cor.clusion 8 of the process.
9 Are there any questions about the process? If 1r not, I would like to turn the floor over to Mr. Lochbaum. 11 MR. LOCHBAUM: Good afternoon. My name is David 12 Lochbaum. I have been the Nuclear Safety Engineer for the i 1 13 Union of Concerned Scient1.'n for UCS since October of 14 1996. Prior to joining UCS I worked in the nuclear industry (3 () 15 for more than 17 years, primarily as a consultant. As a l 16 consultant I worked on assignments at the Perry plant in 17 1995 and again in 1996 For one of these assignments I 18 develooed a lesson plan on design and licensing basis 19 requirements and presented it to managers, supervisors and 20 staff in the design engineering department. 21 I was a reactor engineer for a total of eight 22 years at the Hatch, Browns Ferry, Grand Gulf and Hope Creek 23 nuclear plants, all boiling water reactors similar in design 24 to the River Bend and Perry plants. Among other reactor 25 engineering duties, I was responsible for the fuel integrity ()
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ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0934
14 l l 1 monitoring program and updating the accident analyses for , 2 each operating cycle. i ]%/ 3 I am here today because UCS submitted two i 4 petitions to the NRC last year. Last spring we had provided 5 the NRC with our technical report documenting the concerns 6 we had with nuclear power plants that operate with damaged 7 fuel. At that time we were not aware of any plants i 8 operating with damaged fuel but plants that operated with 9 damaged fuel in the past and were likely to operate with 10 damaged fuel in the future. 11 Five months later we learned that the River Bend 12 plant was operating with damaged fuel, leaking radiation. 13 Other than a letter from the NRC acknowledging receipt of 14 our technical report, we had not heard from the agency. We n ( ) 15 didn't know if they agreed with, disagreed with, or even 16 understood the concerns in our report. 17 In an effort to prompt the NRC to take some kind 18 of action on our safety concerns, we submitted a petition 19 asking that the River Bend plant be immediately shut down 20 until the leaking fuel was removed or until the plant's 21 owners had performed an analysis showing that it was safe to 22 operate with the damaged fuel. 23 A few weeks later 7 learned that the Perry was 24 also operating with failed fuel. So we submitted a similar 25 petition. It is our intention to continue filing petitions [] (_,/ ANN RlLEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
15 l 1 when plants operate with leaking fuel until the NRC T -2 addresses our concerns, or until we run out of postage [Q 3 stamps, and we have plenty of stamps. 4' The River Bend plant, located north of Baton.
-5' Rouge, Louisiana, was granted an operating license by the 6- NRC on November 20th, 1985. The Perry plant, located
- 7. northeast of Cleveland, Ohio,' was issued an operating- ,
8 ' license about a year later on November 13th, 1986. l
-9 ~In each case, the NRC issued the operating license {
10 after a lengthy, deliberative process, through which it 11 concluded that there was reasonable assurance that two 12 criteria were met. The first criteria was that the 13 facility's design met.all applicable regulatory 14 requirements. The second criterion was that the facility q f 15
) would be operated and maintained in accordance with all 16 applicable regulatory requirements. l 17 By letters dated September 25th, 1998 and November 18 .9th, 1998, UCS petitioned the'NRC to require the immediate 19 shutdown of the River Bend and Perry plants because they 20 continue to operate even though some of their nuclear fuel 21 is-leaking radioactivity, a condition that violates the 22 terms of their operating license. As a result of this 23 illegal activity, plant. workers face greater risk from 24 radiation overexposure during normal plant operation. In j 25 addition, both plant workers and the general public may face ;h ANN RILEY & ASSOCIATES, LTD.
V Court Reporters 1025 Connecticut Avenue, NW, Suite 1014' Washington, D.C. 20036 (202) 842-0034
p , i 16 1
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1 -greater risk from radiation everexposure if an accident were-
'2/
to occur. 3 The NRC elected not to shut down the plants.
- 4. According to an internal NRC document that we obtained via 5 the' Freedom of Information Act, the NRC's rationale in the l 1
61 . Perry-case was, "The 2.206 Petition-Review Board concluded 7 :that no' urgent safety problems were uncovered that would 1 6 warrant shutdown of the plant. The clad damage reported is 9 insignificant and it is allowable provided the reactor 10 coolant chemistry were maintained within -- were within 11- permissible technical specification limits as defined in. i 12' Tech Spec 3.4.8. These limits are set to minimize 13 radiological consequences of a postulated design basis l 14 accident and to meet appropriate acceptance criteria. The vs () 15~ petitioner does not allege that Perry Nuclear Power Plant 16 had operated outside the technical' specifications." 17 As the petitioner, I fully agree that ana never i 18 alleged that' Perry, or River Bend, for that matter, operated 19 outside the technical specifications. Hence, the NRC's 20 conclusion seems to have been based on things that we did 21 not say, but.it is very disturbing that the NRC failed to 22 address the things that we did say in our petitions. 23 In the River Bend petition, we specifically 24 referenced 10 separate sections within the plant's updated 25 : final safety analysis report that we felt were being
/~'\ ANN RILEY & ASSOCIATES, LTD.
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17 1 violated by operation with failed fuel. In the attachment [ 2 to our Perry petition, we referenced more than a dozen
.3 separate Updated 2inal Safety Analysis Report sections, but ,
_.4 we didn't reference a single technical specification 5 sectian. The documentation'that we obtained from the NRC's 6 evaluation does not contain a single word about these UFSAR 7 violations. 8 The technical specifications are a part of the 9 operating license. They define conditions that must be 10 satisfied for the plant to operate. Plant ov*ners can change 11 the technical specifications only after formal review and 12 approval by the NRC. 13 The Updated Final Safety Analysis Report, or 14 UFSAR, describes the plant's design features and the safety p-Q 15 analyses that are performed of the plant's response to 16 postulated accidents. It also defines the conditions that i 17 must be satisfied for the plant to operate. The UFSAR is i 18 the primary document reviewed by the NRC in reaching the ' 19 decision to grant an operating license. The NRC failed to
-20 even look at this key document when it evaluated our 21 petitions.
22 We feel it is wrong to evaluate this concern
'23 solely on the basis of the technical specification 24 compliance. More importantly, the NRC also knows that it is
'25- wrong. I call your attention to a notice send by the NRC to O
D ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
1 all plant owners in March of 1996. "The United States [~)D 2 Nuclear Regulatory Commission, NRC, is issuing this 3 Information Notice to alert addresses to instances of 4 reactor operation that may not conform to the licensing 1 5 basis. It is expected that recipients will review the { 6 information for applicability to the facilities and consider 7 actions as appropriate to avoid similar problems. On August I l 8 21st, 1995, the NRC received a petition under 10 CFR 2.206, i 1 9 which was supplemented on August 28th, 1995, that request 10 NRC to shut down Millstone Unit 1 and take enf.orcement 11 action based on alleged violations of license activities 12 related to operation of the spent fuel pool cooling systems 13 and refueling practices. The follow up on the issues raised 14 in the 2.206 petition, including the findings from t (h) 15 investigations conducted by the Office of the Inspector l 16 l General, found that certain activities at Millstone Unit 1 l 17 may have been conducted in violation of these license 18 requirements and that refueling activities may not have been 19 conducted consistent with the Updated Final Safety Analysis , 20 Report, UFSAR." 21 When the NRC received the Millstone petitions in 22 August 1995, they focused on the technical specification 23 issues, even though the overwhelming majority of the 24 concerns involved design issues and analyses in the UFSAR. 25 During the informal public hearing held in April [ h ANN RILEY & ASSOCIATES, LTD.
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19 1 1996 on the Millstone 2.206 petitioner -- petition, Ernie
- 2. Hadley, the counsel for the petitioner, compared the plant's d(N 3 technical specifications to a driver's license, and the 4 UFSAR to the motor vehicle code. He contended that while 5 the driver's license contains a few restrictions on the 6 driver, such as the need to wear corrective lenses, it 7 basically symbolizes the driver's commitment to follow the S requirements in the motor vehicle code in exchange for the 9 privilege to operate a car. Hadley pointed out that when a 10 driver fails to follow the requirements in the motor vehicle 11 code, the state can suspend the license, revoke it, or 12 modify it such that the driver can operate a vehicle only to 13 and from the-job; 14 Ernie Hadley argued that Millstone's owners 0)
( 15 retained the privilege to operate the nuclear power plant if 16 and only if they conformed the conditions within the 17 technical specifications and the Updated Final Safety 18 Analysis Report. Mr. Hadley had been an attorney for many 19 years. He was not expressing his opinion, his desire, or 20 even his philosophy. He was stating his conviction of what 21- the regulations required. History has clearly demonstrated 22 that Mr. Hadley was absolutely right. The NRC was wrong in 23 1995 to focus exclusively on the technical specifications 24 and they are still wrong to do so today. 25 In response to both the River Bend and Perry
/~ ANN RILEY & ASSOCIATES, LTD.
(_s} Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 l (202) 842-0034 { 1
E; 20 1 ~ petitions, the NRC_ indicated.that continued reactor [} 2 ' operation with leaking fuel was-permissible because the V 3 ' radioactivity levels of the. water being circulated 4 throughout the reactor core remained below the technical
~!E specification limit. 'It is true that the reactor water 6 radioactivity limits remained within the' technical.
7 specification limit's. It is'also true that Annapolis is the
'8 capital 1of Maryland and there are 12 inches in a foot. None 9 of the these truths has any bearing on whether River Bend 10 and the Perry plants can operate safely with leaking fuel..
11 What does matter is whether reactor operation with failed 12 fuel is consistent with As Low As Reasonably Achievable, or 13 AUUU4, requirements for plant workers, and with the input 14 assumptions for the accident analyses for the plants. () '15 We contended in our petitions, and we still 16 maintain,- that there is a disconnect on both counts. Before 17 explaining the two disconnects, I will briefly provide some 18 background on nuclear fuel-design and past experiences with 19' failed fuel. This overview is taken largely from our April 20 1998 report,.which was an attachment to each of the 21' petitions. 22 Nuclear power' plants are powered by fuel pellets 23 roughly.the size and shape of a large pencil. eraser, stacked 24 within 12 to 14 foot long metal tubes, sealed at either end 25- with welded metal caps. A simplified drawing of the fuel ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut. Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
i I 21 1 tube 11s'shown in Figure 1.of our report. The fuel tubes are 2! also called the fuel cladding. Fuel tubes are like the gas 3' tank in a car, when the tank is breached, highly volatile I 4- ~ gasoline can' spill out to threaten the safety of its 5 : passengers and innocent' bystanders. .When fuel tubes are i J
- 6 breached,. highly. radioactive material spills out to threaten
- 7. the safety of the plant workers and.the public. l; 8 All. operating U.S. nuclear power plants use. fuel 9 , assemblies containing. square arrays of fuel tubes.
A 10 typical fuel assembly is illustrated in Figure 2 of our 11' report. According to.the NRC, the fuel design basis ensures ] 12 that, "The fuel is not damaged as a result of normal I 13 operation and anticipated operational occurrences." It does i
- 14. not say that-the fuel will suffer only minor damage. It j
() 15 says that the fuel will not be damaged during normal 16 operation. Thus, the fuel design basis includes the 17 Lexplicit requirement for the-fuel tubes to remain intact 18 during normal plant operation. 19 The' splitting or fissioning of uranium atoms 20 inside the fuel tubes releases energy that heats water, 21 atomic energy that powers the plant. Byproducts of the 22 fission process include radioactive gases and solids. These 23' radioactive materials emit gamma rays, along with. alpha and 24 beta particles that could cause damage to the human body. 25: The fuel tubes contain these radioactive materials. When
/~'A ANN RILEY & ASSOCIATES, LTD. ?%_) Court Reporters 1025' Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 r
b 1 22 1 the tubes break, radioactive materials leak into the fuel ( 2 which -- into the water which cools them. This water is 3 = contained within the reactor vessel and the piping connected
- 4 to it, t
5 If the piping fails, contaminated water spills i
)
6 into the reactor containment building. The reactor vessel 1 I 7 and its piping are' located within a reactor containment l 8 building-which forms the third barrier. Because the reactor L 9 containment' building is not leak-tight, it reduces but does i 10 not eliminate the possibility that this radioactive material 11 will escape. 12 Figure 3 in our report from April 1998 shows a 13 simplified drawing of these three barriers. The fuel tubes 14 are the most important of these three barriers. If the fuel ) (% 15 tubes remain intact, the other two barriers can completely () 16 fail and the public will still be protected. The intact 17 fuel tubes contain the radioactive gases and solids and 18 prevent them from being released to the atmosphere. The 19 public cannot be harmed from a nuclear plant accident in 20 which the fuel tubes remain intact, but the River Bend and 21 Perry plants are operating with this vital barrier already 22 broken. 23 Leaking fuel tubes are detected by increased 24 radioactivity levels in the reactor vessel's liquid and. 25 gaseous releases. Not surprisingly, the radioactivity
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23 1 levels rise significantly when fuel tubes break. The causes 2 and precise. locations of fuel tube failures cannot be
}
3 determined till the plant is shut down and the leaking fuel 4 tubes removed and examined. 5 A few years ago, the owner of the Point Beach 6 _ Nuclear Plant in Wisconsin reported a significant event in 7 which, "the fuel cladding was failed to the extent that fuel 8 pellets could be seen through the hole in the clad. 9 However, no pellets' escaped from the rod." The fuel tube's 10 failure ~was detected when the radioactivity levels of the 11 reactor water rose to a level that was 10 percent of that 12 allowed by the technical specifications. In other words, 13 the plant's technical specifications would have. allowed it 14 to remain operating, if that were true, with up to nine () 15 other similarly damaged fuel tubes. This event suggests 16 that the restrictions on reactor water radioactivity levels 17 are too high to prevent operation with gaping holes in fuel 18 tubes. 19 At the Palisades plant in Michigan, three portions 20 of a broken tube were discovered in different parts of the 21 reactor. One segment nearly five-and-a-half feet long was 22 missing about one-third of its fuel pellets. A second 23 segment about four-and-a-half feet long and a third segment 24 .a foot-and-a-half long appeared to contain all of their fuel 25 pellets. This event is disturbing because it highlights how -O ANN RILEY & ASSOCIATES, LTD. km Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
24 1 fragile _ fuel tubes _can become during_ normal operation. At 2 Palisades, this damaged fuel tube literally fell apart as it 3' was being removed from the reactor core and many of its fuel 4 pellets were lost. 5 I will return to the first of the two disconnects 6 I mentioned earlier. This one involves the' radioactivity -- 7 radiation exposures to plant workers. Nuclear plant owners are required by federal regulations to keep the release of
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8:
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9 radioactive materials As Low As Reasonably Achievable, or 10 ALARA. According to the.NRC, "A plant operating with 0.125 11' percent pinhole fuel cladding defects showed a general 12 five-fold increase in whole body radiation exposure rates in J 13 some areas of the plant when compared to a sister plant with 14 high integrity fuel. Around certain plant systems, the () 15 degraded fuel may elevate radiation exposure rates even 16 more." 17 These sister plants were virtual.ly identical 18 because they were built by the same -- at the same time by 19- the same owner, on the same site. The significant variation 20 of radiation exposure rates was not due to thicker concrete 21 or'other design differences, it was due to the operation 22 with damaged fuel tubes. This NRC evidence is troubling 23 because it shows a.significantly increased risk to nuclear 24 plant workers at a facility operating with just 0.125 25 percent fuel tube failures. Many plants consider it ANN RILEY & ASSOCIATES, LTD.
~ %_/ Court Reporters 1025 Connecti cut Avenue, NW, Suite 1014 Wasnington, D.C. 20036 (202) 842-0034
25 1 permissible'to operate with eight times as many fuel tube (') v-2 failures, or up to 1' percent of the fuel tubes damaged. 3 Holes and cracks in fuel tubes release radioactive 4 material into the reactor water. The water carries them 5 through all -- tosall parts of the plant, contaminating 6 equipment throughout the facility. Workers conducting 7 equipment inspections and maintenance on this equipment 8 receive higher radiation' exposures. Indeed, some plant 9 workers have received radiation doses far greater than 10 allowed by federal regulations. 11 At plants like River' Bend and-Perry, the increased
- 12 radiation means that it might take 10 workers to do a job 13 that normally would be done by five workers, so that no 14 individual worker receives an excessive radiation dose. But q_)' 15 this work force received much more radiation exposure than 16 if the damaged fuel were removed.
17 According to Section 12.1.2.1, general design-18- considerations for ALARA exposures of'the River Bend UFSAR, 19 the general design considerations and methods to be employed 12 0 to maintain in-plant radiation exposure ALARA have two-21 objectives -- (1), minitaizing the amount of time plant 22 personnel spend in radiation areas, (2), minimizing i 23 ' radiation levels in routinely occupied ^ plant areas and in i 24 the_ vicinity of plant equipment expected to require the
'25 attention of plant persor nal ANN RILEY & ASSOCIATES, LTD.
'1%_/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014
-. Washington,- D . C . 20036 (202) 842-0034 1
w 26 > 1 It is a'well1 documented fact that plant operation 1/ f7h , 2 with damaged fuel' tubes significantly increases personnel O 3' exposures. Federal regulations require nuclear' plant owners 14 to keep the release of radioactive materials.As Low As i 5 Reasonably Achievable. Therefore, it seems both an illegal 6 activity and a serious health hazard for nuclear plantsito
'7 continue' operating-with known fuel damage.
8_ <There are precedents for taking this action. lJust 9 -last year the owner of the Limrick Nuclear Plant outside. 10 Philadelphia elected to shut down that plant'for about a 11 week to remove leaking fuel bundles. That utility is 12 committed to operating nuclear power plants within radiation 13 levels ALARA. 14 I turn now to the disconnect between River Bend f 1r and Perry operating.with failed-fuel and the design basis on > 16 _ which the NRC issued them an operating license. First, let 17 me explain our understanding of the relationship between 18 normal plant operating conditions and the associated design 19 basis, using an example. River Bend and Perry have an ASME 20 ECode pressure limit of 1,375 pounds for the reactor vessel. 21_- That doesn't mean that the plants can routinely operate with 22 ; reactor' pressure up to but not exceeding that limit. The 23 plants must be able to accommodate'the largest pressure 24 spike that may occur from postulated transients. 25 -Section 15B.S.2.2, overpressurization protection ANN RILEY.& ASSOCIATES, LTD. 1 s-( Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 v 1
27 1 of Perry's UFSAR explains that concept as follows: "The 2 overpressure protection system must accommodate the most (} 3 ' severe pressurization transient so that the ASME Code limit 4 of 1,375 psig is not exceeded. The main steam isolation 5 valve, MSIV,' closure with secondary scrim-(flux scrim) has L 6: been determined to be the most limiting event for 7 overpressure protection." 8 What this means in English is that when the valves 9 in the pipes that carry the steam from the reactor vessel to 10 the turbine close, the pressure inside the reactor vessel l -11 goes up. According to Table 15.0-1, input parameters, and 12 initial conditions for transients in Perry's UFSAR, the l L 13 input parameters for this analysis -- the overpressure 14 analysis assume that the reactor vessel dome pressure is
.I 15 1,045 pounds. UFSAR Table 15B.S.2-3 reports that the peak i 16 reactor vessel bottom pressure is 1,289 pounds for the I
l postulated transient. The pressure numbers are not important to my 19 point. The process, however, is very important. It 20 features three elements, initial conditions, 1,045 pounds; j 21 the' effects of the change, about 250 pounds; and an end 22 point, 1,289 pounds, which demonstrate that the safety 23 margin to the design limit of 1,375 pounds. These three 24 elements form the link between normal plant operation and 25 accident analyses. As long as the plant operates within the [)\s- ' ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 m
r 28 1 bounds established by the initial conditions, the postulated l~'i 2 accident transient can occur with reasonable assurance that Q) 3 the end point will have at least the minimum safety margins. 4 Containment temperature is another example. Page 5- 6.2-3 of-the River Bend UFSAR states the containment design 6 temperature is 185 degrees Fahrenheit. UFSAR Table 6.2-3, 7 conditions for containment response analyses in the River 8 Bend FSAR states that the accident analyses assume that the 9 containment air temperature at the beginning of the accident 10 is a hundred -- or, excuse me, is 90 degrees Fahrenheit, and 11 that is-rises to 141 degrees Fahrenheit following the 12 accident. ) 13 Again, the three elements are present, initial )
)
14 condition, 90 degrees; the effect of the change, 51 degrees; l () 15 and the end point of 141 degrees, which provides safety 16 margin to the design limit of 185 degrees. 17 So the link between the accident analyses and 18 normal reactor operation allows plants to startup, change 19 power and run as long as they remain within the initial 20 conditions assumed in the analyses. As we read the River 21 Bend and Perry UFSARs, it appears that this link cannot be 22 made unless the plants operate with zero fuel failures.
'23 Commissioner Diaz stated during a recent public 24 meeting that zero defects is an unrealistic standard. I 25 whole-heartedly agree with him. However, the burden is on O
(_/ ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
= . - -
29 1 plant owners to define, and the NRC to approve, a more
--I'\ 2 ' realistic. standard. The burden is not -- I repeat, not on U.
3- the public to accept the NRC ignoring violations of federal 4 safety regulations. 5 Let's take a closer look for the missing link 6 between safety analyses and normal reactor operation with 7 respect to damaged fuel. Consider the recirculation flow
- 8. control failure with increasing flow event as described in 9 Perry UFSAR section 15.4.5: " Failure'of the master 10 controller of neutron flux controller can cause an increase 11 in the core coolant flow rate. Failure within a loop's flow 12 controller can also cause an increase in core coolant flow
- 13. rate."
14 In boiling water reactors like those at Perry.and
/~~ '( Tj 15 -River Bend, increasing the amount of flow through the 16 reactor core causes the power of the reactor to go up. Per 17 Section 15.4.5.5, radiological consequences of the Perry 18 UFSAR for this analyzed event: "An evaluation of the 19 radiological consequences was not made for this event, since 20 no' radioactive material is released from the fuel."
21 Now this statement cannot be true unless the 22 initial conditions assume that there.is no leaking fuel. If 23- damaged fuel is present, radioactive material will be 24 released from the fuel. In fact, more radioactive material 25 will'be released due to this event, because the release rate l O ANN RILEY & ASSOCIATES, LTD.
\~sI Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 ,
(202) 842-0034 i i
( ( _
, 30 1 if primarily dependent on_the power level, and~the reactor 2 power level will go up. .[U]
3 But that's not all. Table 15.0-3, a summary of 4 accidents at the Perry UFSAR, lists the number'of failed
- 5. fuel tubes from the safety analyses performed for various 6- postulated accidents. _There's a table here that lists 7 rod-drop accident; it's less than 770 failed fuel tubes.
8 Steam system pipe break-outside containment has zero failed
- 9. fuel tubes. Feed water line' break, zero failed fuel tubes.
10 And the~ loss-of-coolant accident within'the reactor coolant 11 pressure boundary has zero failed fuel tubes. 1:2 Once again these results would appear flawed 13 unless the reactor is operating with no damaged fuel when. 14 the accident begins. In the control-rod-drop accident () -15 analysis, fewer than 770 fuel tubes are calculated to fail 16~ as a result of that event. But are the plant workers and 17 the general public still protected if the accident occurs 18- when there are already three or five or ten, a dozen, or 100 l { 19 leaking fuel tubes? Will-there be adequate protection if 20 nearly 770 fuel-tube failures are added to numerous 21 preexisting leaking tubes? I don't know, and the plain l 22 truth is that no one knows. 23 . Perry UFSAR section 15.4.2.1.1, identification of 24 causes,. describes the control-rod withdrawal event as "the 25' Rod' Withdrawal Error, RWE, transient results from a l I I l
, ANN RILEY & ASSOCIATES, LTD.
Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 l l L_ l
I, o 31 l~ procedura11 error by the. operator in which a single control
, 2 rod or a gang of control' rods is withdrawn continuously A.
3- until the Rod Withdrawal Limiter, RWL, function of the Rod 4' Control and Information' System, RCNIS, blocks further 5- vithdrawal." 6 In boiling water reactors like those at River Bend 1 71 and Perry, withdrawing control rods uncovers portions of 8 fuel tubes, causing'their power output to increase 9 significantly. This local power increase effect may or may 10s not-cause the overall reactor core power level to-increase,
-11 but the power level of the uncovered fuel tubes can increase 12 b'y a factor of 10. Remember that power level is the primary 13 factor controlling how much radioactivity escapes through 14 holes and cracks in fuel tubes. Raising the local power
() 15 level by a factor of 10 means-that much more radioactivity
'16 will be released.
-17 How do the safety analyses-account for this fact?
18 Section 15.4.2.5, radiological consequences of the Perry 19 UFSAR for this event, states, "An evaluation of the 20 radiological consequences was not made-for this event, since
.21 no radioactive material ~is released from the fuel." The 22 same section'of the River Bend UFSAR -- the same section 23' number of the River Bend UFSAR -- contains this statement 24 word <for word, except the comma is missing. They are both 25 correct, but only when this event is initiated with zero l
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ANN RILEY & ASSOCIATES, LTD. h /J . Court Reporters j 1025 Connecticut Avenue, NW, Suite 1014 Washington' D.C. 20036 1 (202) 842-0034 l _z J
32 El' . damaged fuel in the reactor core. When there is leaking (} 2 3 fuel, as there-is at the moment, the-results for the safety analyses for this event are-not applicable.
- 4. Federal regulations do not: allow nuclear plants to 5~ operate with nonapplicable safety analyses. The NRC might, 6 .but-the regulations do not. But. don't take my word for it.
~
7 Look at Table 15A-2-1, unacceptable consequences criteria, 8 plant event category normal operation in the River Bend 9 UFSAR. According.to River Bend's UFSAR, an unacceptable 10 consequence during normal operation is " existence of a plant
- 11. condition not considered by plant safety analyses."
12 As we just reviewed, River Bend's safety analysis 13 for the rod-withdrawal error event considers the fuel tubes 14- to be completely intact before, during, and after this O) ( 15 event. The plant is currently not in the condition covered j 16L by this safety analysis. Neither is Perry. 17 There are other unanalyzed consequences of having 18 an accident with preexisting leaking fuel. For example, the 19' fuel tubes are filled with helium prior to their being
-20~ sealed. Helium is used because of its high thermal 21 conductivity. The leakage of helium through the holes and -)
22 cracks.in the fuel tubes may slow down the transfer of heat 23 .from the fuel pellets to the water. When this heat cannot I 24 be dissipated as quickly as assumed, the fuel temperature i 25 will increase, and may reach the point at which it begins to 'O . ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
33 1> melt the cladding. The leakage of helium from a fuel tube "2 'may reduce heat-transfer rates, .thus potentially increasing f" 3- the chances'that the fuel will'be seriously damaged during a-
.4 loss-of-coolant accident.
- 5. How-much margin is currently available? According 6 .to. Table 15B;6.3-1.of the Perry UFSAR, "The maximum fuel 7 peak clad' temperature for the GE 10 fuel is 2149 degrees F 8 plus the 6 F adders penalties, resulting in a' total of 2155 9- 'egrees d F. 'The maximum-fuel peak clad temperature-for the 10 . cycle 7 GE'.11 fuel was 2184 degrees F,-plus the 6 degree F 11: adders penalties, resulting in a total of 2,190 degrees 12 . Fahrenheit. The maximum _ fuel peak clad temperature for the i
13 . cycle 7 GE 12 fuel is 2181_ degrees F, plus the 6 degrees F 14 adders penalties, resulting in a total of 2,187 degrees F." 15 The maximum fuel peak clad temperature allowed by
'16 Federal regulations is 2,200 degrees Fahrenheit. Thus, the 17 margin at Perry is 10 to 45 degrees for the fuel types.
18 -Will the loss of helium gas reduce the margin by a degree, 19 10 degrees, 20 degrees, or 50 degrees? I don't know. From-20' our research it appears that the NRC and the plant owners 21 don't know either. 2 2 '- What do we know about reactors that operate with
' leaking fuel? We know that the holes and cracks in the fuel 24 tube allow radioactivity and helium gas to leak out. They 25 also allow water to leak in. The high temperature produced
- O ' ANN RILEY & ASSOCIATES, LTD.
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l 34 1 by an operating reactor core disassociates the water into. J -[~ 3 -2 its' hydrogen and oxygen gas constituents. The hydrogen' gas V-3 interacts with the metal fuel tube to form blisters. The 4- blisters embrittle the metal, leading to perforations. 5 Thus, water leaking into a fuel tube may increase the 6 probability that it suffers this type of damage. 7 As a matter of fact, fuel propagation due to this 8- cause has already been identified. In 1993, a fuel tube at )
)
9 the Perry plant experienced a crack measuring 20 inches 10 long, or nearly 13 percent of the fuel tube's length, caused 11 by this water intrusion sequence. So it is known that small 12 holes and cracks can propagate during normal plant operation 13- into rather large cracks. 14 When you release the inlet of a balloon, it moves. A) t 15 Air rushes out of the inlet in one direction, but the 16 balloon travels in the opposite direction. It's a basic law 17 of physics. For every action, there is an equal and i 18 opposite reaction. I 19 If one of the pipes connected to the reactor ' 20 vessel breaks, a classic accident scenario, water and steam 21 would rush out of the opening. The forces inside the 22 reactor vessel are much larger than during normal plant 23 operation. In addition, these forces are in different 24 direction than are experienced during normal operation. 25' What happens when metal tubes weakened by holes
~
ANN RILEY & ASSOCIATES, LTD, (, T Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
35 1' and. cracks are exposed to larger forces from new directions? ( 2 .Recallfthe fragile fuel tube at the Palisades plant that .v-3 . literally fell apart when it was picked up. 4 Does this evidence suggest that damaged fuel tubes 5 can withstand the forces inside a reactor vessel,during an f i' 16 accident?- At the risk of sounding like a broken record or 7 appearing stupid, I don't know. From our research it ) 8 appears that the NRC and the plant owners don't know either, 9- In responding to our petitions, the NRC indicated 10' plants did not need to be immediately shut down because they 11 Lwere within the technical specification limit under 12 radioactivity levels of the water flowing through the 13 reactor core. As I have stated,.that is true. It is also
- 14. irrelevant', because there is no link between a plant
/N t ) 15 . operating just below that technical specification limit and 16 its accident analyses.
17 The technical specification limit corresponds l 18 roughly to 1 percent of the fuel tubes being damaged. Some 19 of the accident analyses, assume that the plant is initially 20 operating with 1 percent of the fuel-tubes damagt" for 21 offsite dose calculations.. But most of these ana yses 22 assume that none of the fuel tubes are damaged when the 23- -accident begins. The analyses assume that nc failures -- 24 ~t he' analyses that assume no failures are clearly not 25 bounding when the plant operates with failed fuel tubes. h%/ ANN RILEY & ASSOCIATES, LTD. Court Reporters
'1025' Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) '842-0034
7.; 36 1 However, even the accident analyses that assume 1
.['T 1 (/-
2' percent'of'.the fuel' tubes are damaged when the accident 3 .' starts are also not bounding. As I have covered, 4 preexisting fuel-tube failures could propagate as a direct 5 result of an accident. Therefore_, compliance with the
'6. technical specification limit does not mean that the workers 7 and the public will be protected from excessive radiation
-8 exposures in event of an accident.
9 Conformance with the design basis requirements of
'10 the UFSAR is needed to assure adequate protection. That's 11 what they're for. Plant operation with damaged fuel is not 12 properly addressed in the River Bend and Perry UFSAR's.
13 There's not-a single word, picture, chart, or table { 14 describing what happens when helium gas leaks out through a 15 hole or crack or what happens when water leaks in. Nothing. 16 In the report we submitted to the NRC last April, 17 I documented a safety evaluation performed per 10 CFR 50.59. 18.' This Federal regulation controls whether plant owners can 19- change how they operate _their plants without prior NRC 20 approval. 21 Prior to joining UCS, I prepared and reviewed 4 22 literally thousands of 10 CFR 50.59 safety evaluations. The ] 23 10 CFR 50.59 safety evaluation that I performed for plants 24 operating with damaged fuel clearly established that NRC 25 approval is required. Neither River Bend nor Perry have ANN RILEY & ASSOCIATES, LTD.
,Q'w/ Court Reporters
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37 f1' sought or been given permission by the NRC. That's a
/ 2 ' vio.1ation of Federal regulations that were created to 3 protect public health and safety.
4' Although I haven't yet heard it, the NRC and/or 5 the. plant owners may argue that the results from the 6 analysis performed for the-classic loss-of-coolant accident sequence-bounds all other accident consequences.
~
7 If.made,
- 8. that argument would be fallacious. It is true that if that
'9 accident were to occur, the presence of preexisting leaking 10 fuel would not make a discernible difference in the amount l 11 of radioactivity released to the atmosphere.
12 For example, a dozen or two damaged fuel bundles 13 at Three Mile Island would not have changed the millions of 14 curies that were -- of reactivity that were released from 15_ the reactor core when it melted down. But there are other 16- accident sequences that are not' bound by this classic 17 analysis. At many plants the maximum radiological threat to 18 control room operators comes from either the control rod l 19 drop accident or the break of a steam-line outside the l '20 ' containment building. The presence of a dozen or two l 21 leaking fu'el bundles, when these accidents occur, could mean
.22 that control room operators receive far more radiation than 23 permitted by federal regulations.
24~ From published news accounts, the NRC and utility 25 representatives have stated that River Bend and Perry are ANN RILEY &' ASSOCIATES, LTD. v( Court Reporters l 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 , I ! 1
1 38
- 1. -not the only plants to operate w_ h damaged fuel; that 2 plant s have been operating with damaged fuel for many years.
3 Their implication is that the history demonstrates that such 4 operation is safe. That is simply.not true. It could be 5 luck and not safety margins that are accounting for this 6 record. i l 7 There hasn't been_an accident at a plant operating
- 8. with damaged fuel, so experience does not demonstrate that 9 plants opurating with damaged fuel are safe, and the safety 10 analyses do not demonstrate that plants operating with 11 damaged fuel are safe.
12 The obvious question, to us at least, is, 13 therefore, why are these plants operating with damaged fuel? 14 Earlier I compared the NRC actions on our petitions to their 15 actions on the Millstone petition. In that case also the 16 NRC and the plant owner argued that what Millstone was doing 17 -- in that case, offloading the entire reactor core to the 18 spent fuel pool every refueling outage -- was okay because 19 everyone was doing it. 20 After further review, it was learned that many of 21 the plants that were doing it should not have been doing it 22 because they had not taken the proper precautions. Had an 23 accident occurred at those plants, they might not have been 24 able to prevent radiation releases that impacted public
. 25 health. They were lucky. Luck is not an acceptable I' ANN RILEY & ASSOCIATES, LTD. \ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
F l i 39 , 1 standard for public protection. i Ii d 2 In our petitions we asked the NRC to order the 3 River Bend and Perry plants to immediately shut down. We s' did.not ask the NRC to take away their keys. Instead, we ; l 5 asked the.NRC to prevent the plants from restarting until 6 .the damaged fuel had been replaced by non-leaking fuel, or 7 until the plant owners had performed safety analyses which 8 demonstrated workers and the public would be protected if an 9 accident were to occur.with pre-existing leaking fuel. That 10 is what the regulations require. We simply asked the NRC to 11- stop being a spectator and start being a regulator. 12 The situation at River Bend and Perry can be I
.13 compared tx) health insurance. When an insurance company '
14 finds out that one of its customers had a pre-existing 15 health problem'that was not accurately reported on the
- 16. medical survey, it will either raise the premium or cancel 17 the policy. It will take this action because the risk 18 factor is higher than the pre-existing condition.
i 19 The difference between this example and the l 20' . situation at River Bend and Perry is that the insurance 21 company will not overlook the matter if the customer 22 correctly spelled all the words on the survey. The 23 ' insurance company, unlike the NRC, focuses on substance. 24 .If the NRC carefully reviews the facts in this 25= natter, we are confident that they will take the necessary
-( ANN RILEY & ASSOCIATES, LTD.
(_ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 r
40 1- actions to protectLpublic health and safety. If the NRC
/'~\ 2 disagrees with our contentions, we would ask that they V
3 document how each one of the many UFSAR discrepancies that 4 we have identified in our petitions and during this 5 presentation are satisfied. We will not accept an NRC 6 denial of our petition based exclusively on the fact that 7 the plants comply with their technical specifications. 8 The 2.206 petition process does not currently have 9 an effective appeal process, so our appeal, like in the 10 Millstone case, will be to the media, to elected officials, 11 and to the public. I have described several safety { 12 questions that must be answered before anybody can 13 truthfully say that nuclear plants operating with damaged j 14 fuel do not pose a threat to workers and to the general O) (, 15 public. At this point no one knows. Ignorance may be { 16 bliss,.but bliss is not safe. What is needed is a bliss 17 reduction program. 18 Thank you. ! 19 MS. ADENSAM: Thank you, Mr. Lochbaum. I'd like 20 now to ask if the NRC Staff has any questions or 21 clarification they would like to ask Mr. Lochbaum. 22' MR. DONOGHUE: Hi, this is Joe Donoghue from the 23 Tech Staff. The only question I have of Mr. Lochbaum is I 24 did.n't see it in your statement today, nor in the April 25 report. Do you have any information or know of any
/\ ANN RILEY & ASSOCIATES, LTD.
~/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
41 11 experimental information that shows a link between the 2 failures or the leaks'that you talk about-and accident ' b%
-3 consequences? That's the essential crux of your argument, 4 and that kind of information'would be really helpful in our 5 evaluation. I 6 MR. LOCHBAUM: I spent literally several months, 7 not continuously, but over a several-month period I looked 8 through the PDR and I have a stack of documents a couple 9- feet tall on everything I could find-about failed fuel. The 10 closest I could come to answering that question is'some work i
11 that was done at one of the national labs, and I forget -- I 12 think it was Sandia, but I may not have the right national
~ l 13 lah, for the Clinch River, case of failed fuel at Clinch !
14 River. Those results may or may not apply to the case we () 15 16 have on lightwater reactors, but that was the only thing I could find anywhere close to answering that question. 17 MR. DONOGHUE: All right, thank you. 18 MR. HANNON: This is John Hannon. On page 9 of 19 10, Mr. Lochbaum, you indicated you had done numerous 50.59 20 safety evaluations for other plants that had been opercting 21 with damaged fuel, and concluded that -- evidently concluded 22 'that a' safety question existed. Are you aware of any other 23 plants that have had prior declaration of such an unreviewed 24 safety. question and-requested NRC approval to operate under 25 that condition? [)
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Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
42 1 MR. LOCHBAUM: Well, we specifically -- first I
'2 . want to clarifyJthe statement'I made here was that I ; 3.l performed a safety evalut. tion generically for -- that was 4 included in our April-1998 report for plants. It wasn't any 5 specific plant, it;was just.a. generic basis. I never
- 6. performed a.50.59 safet.y' evaluation for any plant, specific 7' plant operating.
8L But in answer to the second -- to your question, ' 9 the research'I talked about earlier, we specifically looked 10 on the docket for any requests by any plant licensee for 11 permission to ope.*. ate wit.n f ailed fuel, because that would 12 have set a precedent one way or the other. .We didn't find 13 any, so we assumed that it never was submitted, i 14 MR. HANNON: Thank you. 15 N', CARUSO: This is Mr. Caruso. Mr. Lochbaum, on 16 page 9 of your. statement, you talk about how the effect of a 17 leaking fuel bundle might increase the radiation dose to 18 nontrol room operatm ti. Do you have any analyses which 1 19 evaluate or quantify the effect of the leaking fuel bundles 20 on a control room operator's dose? 21 MR. LOCHBAUM: In my past life I reviewed that 22 analysis for the Fitzpatrick plant, that's one of those 23 design basis documents up there,'and I don't know the 24 perturbation between what was done and what would be done if 25 .you had leaking fuel since I don't think that was -- our ANN.RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
= ,-
43
, 1 contention is that it is not bound by those analyses. I
- 2. think the reason that --
the basis for that conclusion, if 3 you look in the FSAR sections at River Bend and Perry, they 4 assume those accidents -- there's no reactor material
~5 released from the fuel or the steam line break outside 6 containment, and the situation that's there now, that's 7 simply not a true -- would not be a true statement.
8 MR. HANNON: On page 8, when you talk about the 9 leakage of water into operating fuel, do you have any 10 analyses which calculate the amount of water that leaks into
'11 a fuel element during -- that has been damaged?
1:2 MR. LOCPSAUM: Well, from the research we did, the 13 mechanism is normally the water leaks in when the plant is 14 shut down, because when the plant is running, there's the () 15 driving forces to keep the water out. The water leaks in 16 only when the plant is shut down and that differential is 17 nonexistent or reduced. And then when the plant starts back 18 up, the water -- any water that's trapped in there, 19 depending on the size of the leak, it either steams out or { 20 it blows out, like it did at Point Beach. It gets out one 21 way or the other. 22 MR. CARUSO: Do you have any documents that 23 identify any plants that actually started up with known fuel 24 failures, they knew that there was a fuel failure, they shut 25 down and allowed the water to leak into the fuel and then 1 ANN RILEY & ASSOCIATES, LTD. \_- Court Reporters 1025 Connecticut Avenue, NW, Suite'1014 Washington, D.C. 20036 (202) 842-0034
E 44
, l' started up again?-
[]
- x_./.
'2 MR. LOCHBAUM: 'No, I'm not aware of any. Also, we 3 didn't look at'that,.so that's --
.4 MR. CARUSO: You didn't look at that. You also 5 talk in your ---in-this section about how water dissociates 6 into hydrogen and oxygen. Do you know if that effect. occurs !
7 on the outside of the fuel, or is that limited to the water 8 that gets.into the fuel? ' 9 MR. LOCHBAUM: My understanding from the Point f
- 10. Beach root _cause analysis is that occurs inside the fuel 11 tube, 12 MR. CARUSO: Does it occur outside the fuel?
T3 MR. LOCHBAUM: My understanding would be it's 14' outside the fuel. Q. (j. 15' MR. CARUSO: Okay. That's all the questions I 16- have. 17 M
.S. ADENSAM: Are there any other Staff questions?
18 If not, then I would like to go on to Energy Oper:Ations, 19 Incorporated -- 20 MR. LOCHBAUM: Before we do that -- 21 MS. ADENSAM: Excuse me. Certainly. 22 MR. LOCHBAUM: There's -- I know this is unusual, 23 but I have a plane to catch. I have to leave like right 24 now. 25 MS. ADENSAM: Oh, dear.
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f i 45
. ;1- MR. LOf"HBAUM : So I will forego any closing n . 2 remarks.
3- MS ADENSAM: Okay. -Well, we thank you for your
.4 input, Mr. Lochbaum.
l
'S MR. LOCHBAUM: And I apolc ;e for this, but I'll L 6 get'the. transcript and what was --
7 MS. ADENSAM: Certainly. Certainly. 8 I still would like to go on to you, Mr. Titus. 9 MR. TITUS: Thank you. We are pleased to have 10 been invited here today to share our views regarding the 11 Union of Concerned Scientists' petition. We have previously 12 .provided detailed technical responses to NRC questions 13 related to the petition. These responses thoroughly address t 14- all of the Union of Concerned Scientists' concerns and ( ( f 15 questions that we have been made aware of today. We hope 16 our discussions today wi11'be of at.ditional value in helping 17 the NRC achieve closure on this petition. 18 I'would like to note that we have shared our 19 presentation with First Energy in order to hopefully reduce 20 duplication and put more information out in the limited 21 amou7t of time available. 22 Mr. Lochbaum -- as an aside, Mr. Lochbaum went 23 ,over in some detail his resume. I could bore you with a lot 24- of details on the extensive experience we have here, but let 25- ^me just say that I have personally 30 year nuclear power L ANN RILEY & ASSOCIATES, LTD. l L A/~')\ ss ~ Court Reporters l 1025. Connecticut Avenue, NW, Suite 1014 i Washington, D.C. 20036 (202) 842-0034 l 1 m ]
m-m t ] 46 1 experience; Mr. King has 19 years; Mr. Bethay 20; you add up
$/') \-s/
2 1all the experience of all the people we have representing us 4
-3 today,'-it's well over 100 years. So we'believe we have a j 4 'significant amount;of expertise in coming to this informal
]
5 hearing today.
~
6 I would now~like to ask Mr. Rick King, our l
- 7. Director of Nuclear Safety and Regulatory Affairs, to add
- 8 some< additional opening. comments, specifically with respect 9 .to the petition itself.
I 10 MR. KING: Thank you, Fred. 11 Well, good afternoon. EOI has reviewed the Union 12 of Concerned Scientists' petition and has reached several j 13 conclusions. Our discussion that you will hear today
- 14. provides an overview of our conclusions as they apply to i
) 15 -River Bend Station.
16 In summary, I would like to briefly state.what 17 conclusions we have reached: 1 18 Number one, from our view of the Union of 19 Concerned Scientists' petition, the issues raised do not 20 have any safety significance or impose a threat to the 21 public health and safety. 22 Two, we have a strong belief that River Bend 23 Station's safe' plant operations is bounded by regulations 24 and a defense-in-depth safety analysis that establishes our l 25 . licensing basis. i T ANN RILEY & ASSOCIATES, LTD. Cs . Court Reporters l 1025' Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 l (202) 842-0034 i
47
-l 1- Three, we are in compliance with the NRC f~).
(G 2 requirements and no violations have occurred. Based on our 3 review and technical response provided on February lith of 4 this year, Ewe respectfully request that the NRC director's 5 decision to deny the UCS 2.206. petition request. The UCS-6 has raised no new issues that warrant-the requested actions. 7 The River Bend Station design and licensing basis l 8 are founded on proven regulatory requirements and have 9 established safety margins in our approach to provide 10 adequate protection and a reasonable assurance to safety. 11 River Bend Station continues to operate in a safe 12 manner within the plant's licensing basis and the applicable 13 regulations. .This continues to ensure we help preserve and 14 protect the public health and safety. ( 10' ( ,/ 15 At this time I would like to introduce Mr. Steve 16 Bethay. Mr. Bethay has 20 years of original design and , 17 licensing basis experience of nuclear plants, and Steve will ; 18 discuss an overview of the River Bend Station position 19 regarding the Union of Concerned Scientists' petition 20 assertion. 21 Steve. 22- MR. BETHAY: And hopefully that works. Is this
.23 working? Okay.
24 Good afternoon. My name is Steve Bethay. I'm the 25' director of licensing for Entergy Operations. P ANN RILEY & ASSOCIATES, LTD. (']
\.-- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 L
48 1- As Mr. Titus. stated, we're at this informal public
~
2 ' hearing between the Petitioner, the Union of Concerned (} 3 Scientists, the NEC staff, and the licensees to express our 4 views on the operation of River Bend Station with minor' fuel 5 cladding defects.
~
6 Above everything, I want to emphasize that EOI --
.7 Entergy; Operations -- maintains a complete and total 8 commitment to nuclear safety, because we believe if we fail 9 at that, nothing else matters. That's our No. 1 priority.
L10 Our job as a licensee is to ensure the protection of the 11 public health and safety remains our top priority while we 12 effectively deal with the wide range of operating issues so 13 that we are a profitable enterprise. My presentation today 14 is focused on' explaining why the existence of minor fuel 15 -(G) cladding defects is not a threat to the p"blic health and 16 safety; ar d why we belie ' 7 .that - the River Bend st ation _s 17- operating today in full compliance with its design and
' 18 - Jicent .ig basis.
19 We're here today at the request of th, NRC to 20 explain in this public setting that we are operating safely 21 in accordance with NRC regulations and within the scope of 22 our design and licensing basis. And that's despite the fact 23 that we have less than 10 defective fuel pins out of over 24 43,000 fuel pins in the reactor. At the end of my 25 presentation you will understand that the conservative [~ ANN RILEY & ASSOCIATES, LTD. ' \-)/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 l
49 1 regulatory requirements are fully adequate and appropriate 2 and that operation with minor fuel cladding defects is f~/]' x_ 3 within the bounds of our licensing analysis for the River 4 Bend station. 5 Additionally, you will understand that River Bend 6 is in full compliance with all operational limits, limits 7 that we are operating well within, and that these limits { 8 provide adequate protection for the public health and safety 9 and for the safety of the workers at River Bend station. 10 Finally, I hope that you will also understand that 11 we have exhaustively reviewed each of the petitioner's ' 12 concerns, and we believe that they are without merit. 13 Accordingly, I'll request that the staff issue a Director's 14 decision pursuant to Section 2.206 denying the UCS petition p) t y 15 for River Bend station in fully, 16 I'd like to go back and just talk a little history ! 17 on design and licensing basis. Historically, the design and 18 - licensing basis for U.S. nuclear powerplants has been based 19- on conservate id defense in depth. The NRC has 20 established c:;r servative regulatory requirements and 21 guidance such as those found in 10 CFR 20, 10 CFR 100, a 22 host of regulatory guides, and the standard review plan. 23 These regulatory limits provide margin to the protection of 24 the public health and safety. 25 As a licensee, we strive to maintain a
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(_/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
y. 50 1 . conservative design and operating philosophy. As 2 responsible engineers e we strive for excellence, and (V ) 3 therefore consider a wide range of possibilities, including 4 worst-case conditions. -Consequently, our facility has
.5 ' defense in depth, just like every other nuclear plant in 6 this country. We have three significant barriers to the 7 release-of radioactive material from the fuel, We have.the 8 fuel cladding. We have the reactor coolant system pressure
- 9. boundary. And we have the primary containment building.
10 Our plant design considers possible imperfections 11 in each of these barriers and confirms that the public and 12 the workers are safe before operating within the bounds of 13 our licensing basis analysis. As part of the design 14 process, a wide range of analyses are performed in'
/3
( ,) 15 accordance with the regulatory requirementi to provide 16 margin to the regulatory limits under~a va2iety of both 17 normal and abnormal operating conditions. j 18 River Bend station is licensed by the NRC to !
\
19 operate within the bounds.of these analyses. These analyses J 20 are performed with conservatisms built in to ensure that the I 21 licensing basis is more conservative than what the I l 22 regulations might allow. 23 Finally, operation of River Bend is conservative l 24 with respect to its operational or technical specification j 25 -limits. We've established procedural requirements and ! l ANN RILEY & ASSOCIATES, LTD. ( e Court Reporters 1025 Connecticut' Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
1 l l 51 I l' management limits that are'much more conservative than those
. f,j'N 2 . which might otherwise be allowed by our operating license.
3 We'rs committed to protect the public, the environment, our 4 employees,_and our investmsnt. 'Our conservative operating l i
'S philosophy adds additional, margin to ensure that we don't
~ 6- . operate'our facility in a manner that challenges our design 7 or our licensing basis. 8 What I've just described may be better illustrated L
)
l [ 9 with this chart which pictorially shews how our l
.10 -defense-in-depth approach relates to conservative daily 11 plant operation. Our operating philosophy within Entergy 12 Operations dictates that we operate f.n the white zone, the 1 13' me agement restrictions area. This is controlled by plant 14 procedures that establish limits morrs conservative than the j fN q ,) 15 values described in our licensing basis documents, including 16- the technical specifications.
17 We operate in'the white zone of the management 18 restrictions area with significant margin to our operating 19 limits even with a small number of defective fuel pins. The 20 operating limits established by the technical specifications 21 and other licensing basis documents are conservatively set 22 to ensure that we're always bounded by our accident 23 analysis. Analytical results include significant 24 conservatism in initial conui' Ac is to ensure that the 25 radiological guideline values vf 10 CFR Part 20 or 10 CFR ANN RILEY & ASSOCIATES, LTD. k- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
b 52 l' Part 100 will not be exceeded. 2 One of the initial conditions considered in the 3 analysis for normal operations and abnormal operating events 4 and accidents, which I'll discuss in a moment, is the 5 possibility for preexisting minor fuel cladding defects. 6 Finally, we also have conservative NRC regulations 7 that ultimately bound our operation. As you are well aware, 8 these regulations are designed to protect the health and 9 safety of the public. Operation in conformance with these 10 regulations is therefore by definition safe plant operation. T 11 The gray, blue, and green regions of this simple 12 chart define the licensing basis of the plant. Our 13 ' operation has not approached any regulatory or safety limit. I 14 It is within the bounds of our analyses. It is within the { p
,j
, 15 bounds of our operating limits. In short, our management'
)
i 16 restrictions bound our current operation and by definition I 17 ensure that we're extremely safe and in compliance with our 3 1 18 licensing basis.
)
19 The petition before us today cited a number of j
- 20. excerpts from the River Bend safety analysis report. These 21 quotes appear to be taken out of context, and when read by 22 themselves can lead to erroneous conclusions, As you well 23' know, the safety analysis report is a summary-level 24 licensing basis document. It provides bounds for the safe 25 ' operating envelope of the plant. The River Bend safety
,O ANN RILEY & ASSOCIATES, LTD. \/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
n 53 1 analysis report is complete, and it defines the licensing f%. 2 basis of the plant quite well. su )
] But you can't just look at 3 isolated sections. It's often necessary to look at several 4 different sections of the FSAR in conjunction or at the same 5 time-in order to gain an accurate picture. The critical 6 question is whether the bounds established in our licensing 7 basis envelope the situation that we're discussing he.
8 .today. As I've stated, our operating limits protect -ei 9 analysis results; 10 CFR 20 and 10 CFR 100, adequately i 10 protect the public health and safety. 11 On February 11, 1999, we responded in writing to ' 12 the UCS petition. That letter covers in detail the basis ' 13 for our belief that the current operational condition of i 14 River Bend station is well within the bounds of the A ( ,) 15 licensing basis of the plant. For this presentation, I just 16 want to point out that the operation with fuel cladding 17 defects is specifically addressed in these licensing basis 18 documents, in BWR design basis documents, in the NRC 19 standard review plan, chapters 11, 12, and 15; in the 20 plant-specific safety analysis report, chapters 4, 11, 12, 21 and 15; and in our technical specifications, specifically in 22 sections 2.1.1, 3.4.8, and 3.7.4. 23 Additionally, radiological evaluations account for 24 the impact of fuel cladding defects during normal operation 25 and anticipated operational occurrences, as well as design []
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.ourt Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
T 54 1 basis. accidents. Specifically, normal operation and O 2' anticipated operational occurrences are addressed in FSAR L)
- 3. chapters 11, 12, and 15. Design basis accidents are 4 addressed.in chapters 6 and 15.
5 River Bend was. licensed in the early 1980s. 6 Clearly the issue of operation with minor fuel cladding l { 7 defects was addressed as part of the original licensing l 8 efforts. 9 The safety analysis report considers two f 10 fundamental operating' states, normal operation and 11 anticipated operational occurrencer and design basis 12- accidents. I want to spend a few minutes discussing the 13 licensing basis for each of these operational states in the 14 context of operation with minor fuel cladding defects. 15 First, normal operations and anticipated 16 operational occurrences or transients. The FSAR chapter 11, 17 and more specifically the nuclear safety operational 18 analysis, or NSOA, that's found in chapter 15 of the FSAR, 19 specifically consider the possibility for preexisting fuel 20 cladding defects. In either case, whether normal operation 21 or during an anticipated operational event, we're bound to 22 comply with the requirements of the technical specifications 23 in 10 CFR Part 20. 24 Operational limits are imposed to ensure that we 25 are within the bounds of our analysis. The bounding limits
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55 for fuel defects are specific activity in the reactor 2 coolant and offgas system pretreatment radiation levels.
/'%LN' - If
-3 plant operation is within these limits, we are assured that, 4 should a transient or an accident -- and I'll talk about
'S that in a minute -- occur, we would not exceed a release 16' outside of what has already.been analyzed. And our 7 management restrictions, like the chart I showed a minute '
8 ago,.are imposed to ensure that our operational limits are 9 not reached. 10 Now just so that you know where we're operating 11 today, our' current status is that we're operating the River 12 Bend station, and the offgas pretreatment radiation levels 13 are about 3 percent of the tech spec limit. The reactor 14- coolant system specific activity is about 5 percent of the f 15 tech spec limit. So we're operating well within what'I 16 characterized as the management rectriction area and the 17 operational limits. And River Bend station is clearly 18 operating in conformance with the NRC requirements and 19 regulations. 20 Moreover, the management restrictions that we've 21 imposed on ourselves would have us consider a shutdown wel.' 22 before we reach what our operating license might otherwise 23 allow. 24 As a result of rigorous adherence to our
'25 administrative limits, we have a very clean plant O)
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56 1 . radiologically, and we plant to keep it that way.
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2' Speaking'a little more specifically about ,
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3- anticipated operational occurrences or transients, the UCS 4- . petition suggests that minor fuel cladding defects have not {
- 5. been adequately considered in evaluating these scenarios. ;
6 We strongly believe that the current NRC-approved methods j , -7 for analyzing anticipated operational occurrences are fully 8 ' adequate. l 9 The purpose of transient analysis is to i
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10 demonstrate that the event by itself doesn't cause any new 1 l 1 11 failures. This is accomplished by showing that during the l 12 event that important reactor parameters such as pressure, 13 temperature level, critical power ratio, those limits are 14 not exceeded. All. operational limits still apply. .The event is simply an extension o'f normal operation.
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15 16 Therefore, the name " anticipated operational occurrence." l l 17 Plant operation following such an event still has to remain 18 in compliance with the technical specifications. It still 19 has to remain in compliance with 10 CFR Part 20. Therefore, 20 the transient results are bounded by normal operations. 21 The petition alleges that we have not adequately 22 considered the effects of anticipated operational 23 occurrences on preexisting fuel-cladding defects. As I've 24 stated, normal operation and anticipated operational 25 . occurrences are the same thing from a regulatory
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N/ _ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
57 1 perspective. That is, the same regulations and operating
,x I
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) 2 limits apply in either case. Whether a minor fuel cladding 3 defect occurs and grows during normal operation, or somehow 4 grows as a result of an anticipated event, the result is the 5 same. We still have to comply with the tech specs in 10 CFR 6 Part 20. The current required action levels bound our 7 operation no matter what may cause an increase in offgas 8 pretreatment radiation levels.
9 And, finally, just for a little historical 10 perspective on transients, there was GE work performed about 11 25 years ago which addressed transients with fuel defects 12 and showed it's highly unlikely that a defect would grow 13 significantly during a transient due to the short duration 14 of the event. That information was submitted to the NRC. /^N ( ,/ 15 Moving on to design basis accidents, realistically 16 there would be minimum fuel cladding damage as a result of 17 an event. We performed extensive analysis for a design 18 basis accident using NRC-approved methods which 19 demonstrate -- and that would be in accordance with 10 CFR 20 50.46 or 10 CFR 50, Appendix K -- and these analyses 21 demonstrate that realistically you probably would see very 22 minimal fuel clad defects. But, regardless of the results 23 of the accident analysis which show little or no fuel 24 damage, our radiological analysis for design basis accidents ; l 25 assumes a significant amount of the fuel in the reactor is I (i ANN RILEY & ASSOCIATES, LTD. 'k / Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
r ' 58 1 damaged. (
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( 2 .Our calculated-offsite doses must not exceed 10 N. l 3 CFR PartL100 guidelines or the appropriate fraction of Part 4 100-guidelines for.that event. Our operation within the 5 limits maintains the validity of these analyses, and clearly , 6 -preexisting fuel defects are bounded by the assumptions of 7 the radiological analysis'for design basis accidents. So I
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8 think it should be clear to you that we have considered 9 preexisting defects. They're considered as a part of a 10 normal operation explicitly in the FSAR that anticipated 11 abnormal events are simply an extension of normal 12 operations, therefore the same limits apply, and the assumptions of the radiological analysis for design basis 1 13 14 accidents clearly bounds an initial condition that assumes a p ( .. 15 few pins with defective cladding. 16 Now I'd like to shift gears and address operation 17 with minor fuel cladding defects as it pertains to l 18' maintaining our radiation exposure levels as low as 19 reasonably achievable. The River Bend ALARA program is 20 fundamentally based on design features which include 21 shielding and effluent processing, dose assessment, which 22 includes personnel monitoring and area surveys, and 23 administrative controls for things such as work planning and 24 access controls. And specifically FSAR chapter 12 discusses l 25 the River Bend ALARA program. I'd also like to point out (T Nl ' ANN RILEY & ASSOCIATES, LTD. Court Reporters i 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 l
IL 59 ! 1 that our current operation with the fuel cladding defects we l- ! l
,/~ 2 have today we've seen no increaseLin radiation dose levels L )) - l
[ l 3 in routinely occupied areas of the plant, nor.have we seen l L i 4 any increase in our actual offsite release levels. l 5 Next slide. For River Bend Station, the original i licensing basis dose assessment, performed during the 6
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7 original licensing days, estimated an average value of 948-8 person-rem per year, and this was considered a pretty good 9 level when the plant was going through licensing 15 years 10- ago. In 1998, the actual River Bend total accumulated dose
.11 was 54 person-rem. Clearly, the ALARA concept is well 12 established at River Bend and further reflects our 13 commitment to'a strong safety culture.
14 Shutting down'to remove fuel that has minor fuel- ; () 15 cladding defects would actually increase our personal -- 16 personnel radiation exposures. River Bend historically 17 operates at approximately .14 person-rem per day during 18 normal operation. 19 We discovered the first fuel cladding defect on 20 September 21st, 1998. That was 194 days prior to the next 21- refueling cutage, which is currently scheduled to begin
- 22. April 3rd. If River Bend were to continue to operate for 23 the 194 days at a dose rate of .14 person-rem per day, the 24 occupational dose would be about 28 person-rem. If we were 25 to shut down to remove the defective fuel bundles, an ANN RILEY & ASSOCIATES, LTD.
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r 60 1 additional dose would occur. If you assume a 14 day outage I v) 2 3 to remove the bundles, and assuming no other dose contribution, that is, no other work going on during the 4 outage, the outage dose would be approximately nine 5 person-rem. That would be in addition to the other 180 days 6 of operation with the fuel cladding defects at about .14-7 person-rem per day, giving a total of about 35 person-rem. 8 Note that in either case, normal operation, with 9 or without fuel cladding defects, is about the same, about 10 .14 person-rem per day. So, clearly, with the number of 11 fuel cladding defects at River Bend, it would be 12 inconsistent with the ALARA principle to shut the reactor at 13 this time for the sole purpose of replacing fuel assemblies. 14 Now, of course, the defective fuel assemblies, all n (s_, ) 15 the identified defective fuel assemblies will be replaced { 16 during the outage in April, which starts in April. 17 So, in conclusion, River Bend Station is being I I 18 operated safely and in compliance with its licensing basis j t 19 and the ALARA principle. We believe that the current ) l 20 regulatory limits and policies are fully adequate. We ! 21 believe operational limits in 10 CFR 20 limit the effect of f 22 potential fuel cladding defect growth during normal 23 operation and abnormal events. We believe that our l 24 conservative radiological analyses for design basis l 25 accidents fully bounds the consequences of accidents with i l (~')
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ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 i Washington, D.C. 20036 (202) 842-0034
P i 61
~ li pre-existing; cladding defects, ' that operation' with fuel .f['h- -2 cladding defects below current limits is not a threat to the
(/ 3 public health and safety. Our. current occupational-4 radiation exposure levels are at historical lows.and.
.5- ; operation with minor-fuel cladding defects is preferable to i
6 a forced shutdown to remove the fuel-from an ALARA l l 7 perspective. 8' The issues raised by the Union of Concerned 1 9 Scientists do not constitute a threat to the public health
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I 10- and safety. The petition and the information supplied at
'11- this meeting have not established that the UCS issues have i
12 any safety or regulatory significance. Based on our (i
- 13 consideration of the applicable regulatory requirements, the
'14 : conservatism of our design and analysis, and our
() 15- conservative operating' philosophy, we conclude that there is 16 no concern that occupational dose limits could be exceeded. 11 7 There is no concern for the potential for significant core 18 damage. There is no reduction in the protection of the 19 public health and safety. And that since the UCS petition j 20 does not raise an issue that is either a violation of NRC 21 requirements, or an' issue of safety significance, we ask 22- that the NRC staff issue a Director's Decision fully denying
'23- :all claims raised by the UCS. And we would be happy to 24 address any questions you may have.
25 MS. ADENSAM: Thank you, Mr. Bethay. Does the NRC i
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<1 . staff haveiany questions?
E 25 MR. DONOGHUE: 'Ifhave'two questions for you. 3 '. .MR;:BETHAY: Yes, sir.
=4: MR;!D7NOGHUE: -This is Joe-Donoghue from the: tech
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'5 istaff. I"think I-am' going. backwards in your presentation. s
- 6' ;You compared the dose-that you would receive from replacing 7 the' damaged fuel.rather.than operating. .I-didn't'see any. _
j t 8 -11th February document -- I didn't see those" numbers.in
-9' -there. Did I miss that, or are you going to provide that 110 documentation for us? That would be helpful for us to see.
'11 MR. BETHAY: I don't believe'- ~that was not i
~12' specifically included in the, submittal.
13 MR. KING: If I may, I believe.there were some 14 .overall. conclusions relative to that, but the specific .
- 15 numbers may-not have actually.been in there.
16 .MR. DONOGHUE: Okay. Well, having the numbers'to 17- back-up the statement would help us. That is one point.
'18 MR. BETHAY: Okay.
119 MR. DONOGHUE: Another question is related to the 20; '11 February document. On page 9 it discusses some 21 . experience with failed fuel rods or at least some testing,
- 22 reactivity assertion testing that was done with hydrited -
23 a fuel with a hydrited-cladding, but there is no reference
- 24. cited there. I am aware of some testing that sounds like 25 this, butcI would be interested in' knowing what the specific T h. ANN RILEY.& ASSOCIATES, LTD.
O Court Reporters 1025 Connecticut. Avenue, NW, Suite 1014 Washington, D.C. 20036~ (202) 842 0034
t i 63 i test is that you are talking about to make sure that we are n 2- -- I understand what you are talking about. .(/-).
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3 MR. BETHAY: Do you have that reference? 4 MR. KING: I don't have that' specific reference 5 with me, but we can provide that to the staff. We do have 6 that have.that '.ere, so we can get to you before the meeting i i i 7 -ends. { 1 8 MR. DONOGHUE: Thank you. Thank you. l l 9 MR. HANNON: This is John Hannon. I just wanted
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10' to get a clarification on your point about the occupational 11 exposures. If I understand what you said, it is that you 12 are not seeing any exposure increase due to these minor 13 number of pin leaks, pinhole leaks. Is that due to 14 intervention? Are you doing shielding or doing some shorter () 15 stay times, or are you taking any -- are you changing your 16 practices any to achieve a net stable dose rate? 17 MR. BETHAY: I don't believe we have changed any 18 practices, or we haven't changed any policies or procedures. 19 .There are still area surveys done. There is still 20 monitoring done. It is my understanding from the people 21 that are actually in the Health Physics Department, that 22 they see some increased dose levels in very isolated areas 23 of the plant, in the offgas system, but the general areas, I 24 routinely occupied areas of the. plant are seeing essentially 25 virtually'no increase in dose levels, so we are not taking (~') ANN RILEY & ASSOCIATES, LTD.
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v:, 64 1- any'particular actions, no. L M 2 MR '. HANNON: Thank you. Q, 3 MR. CARUSO: In.your calculated comparison of dose
. _4 rates.of accumulated doses,*ihere you compare the dose that you would accumulate from shutting down versus not shutting
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5 6 down, did you consider dose rates into the future that are 7 going to accrue because you do not show down now and, 8l 'therefore,1ycu accumulate actis..ty at places in the system < 9 that'you will have to deal with into the future? 10 MR. BETHAY: Let me restate your question so I 11 understand what you are asking. I assume you are asking, 12 did my numbers consider played out or whatnot in systems it. 13 may affect.for some time down the road? 14 MR. CARUSO: yes. (f 15 MR. BETHAY: The short answer is no, sir. I was 16 looking specifically at the period from September 21st to 17' the. outage in April. 18 MS. ADENSAM: Mr. Bethay, I have one. question, if
'19 I may. When you talk -- what I heard you say, that you 20 calculated nine person-rem to remove the damaged fuel from 21 the reactor vessel, is that correct?
22 MR. BETHAY: That's correct. 23 MS. ADENSAM: Is that exposure accrued even if you 24 waited until the outage? .I mean if you wait until your 25J refueling outage to remove the fuel, would you not -- would [}
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ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut' Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
65 j 1 'you' calculate a'different numberEto remove this damaged 'l 2 fuel?
'3. -MR. BETHAY: It would probably be.a similar
'4 number. The point is we don't want to-accrue that dose 5 twice, once for a'short-term outage specifically for the 6 purpose for removing fuel, and, again, during a normal I 7 refueling outage when we will'have other in-vessel work,
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8 other refueling. activities going on. l 9 -MS. ADENSAM: Okay. I guess I am trying to
'1 10 understand, would that be additional during your outage 11 because you have' damaged fuel, or would that be part of what 12 _you'would normally expect to see? l 13 MR. BETHAY: That would-be a number that you would ;
- 14. normally expect to see during a refueling outage for D~
( ). 15 refueling floor type activities. i 16 MS. ADENSAM: Do I understand then that you do not 17 anticipate any additional personnel' exposure as a result of 18 having to remove.the damaged fuel? 19 MR. BETRAY: That's right. HWe don't anticipate 20 additional dose beyond what you would normally expect during 21 the. outage to remove the fuel. 22- MS. ADENSAM: Thank you. Are there any other 23 staff questions?- 24 [No response,] 25 [*S . ADENSAM : If not,'since.we seem to be a little fI ANN RILEY & ASSOCIATES, LTD. i/- s- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
r-66 i 1 bit ahead of time, I think we could afford maybe a 10 minute j
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[') 2 break, but we must start back here promptly at 25 of 3:00, w/ ' 3 please. 1 4 (Recess.] I 5 MS. ADENSAM: I would like to ask everyone to get 6 seated. If we could get started again, please. I 7 understood that EOI had the reference, and I thought I would l 8 give them an opportunity to read that into the record if ! 9 they would. 10 MR. KING: Yes, Ms. Adensam. Let me follow up on 11 a question that was asked by the NRC earlier, at the end of 12 the River Bend's presentation. You had mentioned that on 13 page 9 of our February lith, 1999 submittal, there had been 14 reference to some testing and a reference for that, and that l 15 reference is a document that was presented at the 1997 [/ x- i 16 American Nuclear Society International Topical Meeting, 17 entitled, " Regulatory Assessment of Test Data for Reactivity 18 Accidents," by Mr. R.O. Meyer, and that was held in , 19 Portland, Oregon, March 1997, ANS meeting, so we can provide i 20 that to you if you would want it. 21 MR. DONOGHUE: Thank you. 22 MS. ADENSAM: Thank you, Mr. King. I would like 23 to turn the floor over now to Mr. Lew Myers of First Energy. 24 MR. MYERS: Good afternoon, I am Lew Myers, the 25 Vice President of the Perry Nuclear Power Plant. I would [~')
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E 67
- 1. like to personally thank the Nuclear Regulatory Commission L
() 2 and.the staff, as.well as-the local media and the people 3- back at Perry for the opportunity to come here today and l 4 . provide our perspective on the plant operations, as well as l 5 our perspective on the fuel performance regarding this
- . 6~ issue.
7 Sincethe defect first occurred, I have had six I 8'- management focus areas that I would like to discuss with you 9 today. The first'one is safety. Safety has been and will 10- continue to be the top priority of the Perry Nuclear Power 11- Plant. We have been and we will continue to provide -- I'm 12 sorry -- we will continue to operate the plant well within 13 the licensing limits and the design basis.
.14- At Perry, what we are talking about today is three
() 15 potential minor fuel defects, not failures, and over 120 16 miles'of fuel rod. We suppress these-fuel rods to prevent 17- further degradation and there had been no impact on safety 18 or health to any of our employees. We are presently 19 operating the plant at point-eight percent -- ladies and
-20 gentlemen, that is .08 of our licensing limit on reactor )
-21 ' coolant system chemistry.
12 2 What does this mean? It means that the : 23 . degradations could increase significantly prior to us seeing j 24 any of'our limits, approximately 100 times. 25; My second area of focus has been the community i I y ANN RILEY & ASSOCIATES, LTD. s_/. Court Reporters
.1025 Connecticut Avenue, NW, Suite 1014 i Washington, D.C. 20036 l (202) 842-0034 p
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g 68 1 relations. We immediately notified the NRC, as well as the lT~ 2 local news media, when we found the fuel performance
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3 problems. We wanted to demonstrate to the community that we 4 were operating the plant both effectively and safely. The
- 5. community members.put their trust in us to operate this 1
6 . plant safely, and we must provide them with accurate i l 7 lifetime information. I want to tell you, I take this 8 . responsibility very seriously. j 9 My third area of focus has been our continuing 10 operational capability. I have brought our Operations , t 11 Manager with us today, that I introduced earlier. We are 12 committed to maintaining the Perry plant in good material 13 condition, it is in good material condition. The pinhole j 14 defects have not_ impacted our routine daily operations. The () 15' plant continues to operate near capacity and is operating 16 well within the NRC regulations. There are no potential -- 17 there is no -- if there were any potential for the pinhole 18 defects co increase and effect our routine, daily, safe, 19 reliable operations, I would shut this plant down. In fact, 20 'we have already created administrative guidelines that would 21 shut the plant down well before exceeding any of our 22 operational limits. 23 My fourth focus area has been the impact on our 24 employees. I am pleased to stand here today and to tell 25 everyone that there is no increases in radiation at the
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69 1 . Perry Nuclear Power Plant. The fuel defects have been 2- suppressed and we are seeing no increases in either ( ). 3 radiation or contamination within our plant. In fact, Perry 4 -had one of the lowest doses of any boiling water reactor.in 5 history in 1998,.39 rems. We are committed to keeping the 6 ; plant material condition good and keeping the doses As Low-7 as Reasonably Achievable.
- 8. I took the Union of Concerned Scientists' 9 recommendations very seriously. We evaluated the Union of-10 Concerned Scientists' recommendation to shut the plant down 11 and replace the degraded fuel rods with lots of our 12 engineers. You know, we have.some very good engineers, too.
13 If we replaced the fuel rods now, we would require two 14 refueling outages, two shutdowns. Our employees would () 15' receive significant additional exposure, about as much as.
- 16. they typically receive in the whole year. I don't believe -l 17- this action would demonstrate As Low As Reasonable- ,
18 Achievable principles that is our.present license. I 1
- 19. My fifth focus area has been the refueling outage.
20 We will replace the defective rods during the: upcoming i 21' refueling outage. We are also committed to inspecting all 22~- of our fuel for potential defects in our normal routine j 23 outage on-March 26. 24 My final. focus area, maybe the most important for 25 -me, is the gaseous effluents. That is because that is you, (T A/ ANN RILEY & ASSOCIATES, LTD. w Court Reporters
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70 1 the public. There has been no increases in radiation inside (g - j 2 our plant or to the environment, no increases. It is
.3 important because my wife andlI, as well as most of our l 1
4 employees, live in'the local community. I can see the I i 5; cooling towers from my house. Many of our employees, 800 l 6 employees, send their kids to public schools three-quarters 7 of a mile away from our plant. 8 -I want to end this by emphasizing today that there J 9 has been no impact on safety or health to our employees or 10 to any members of this community. Thank you. With that, I 11 would like to turn the presentation over to Howard 12 Bergendahl. Howard is our Director of Nuclear Services, he 13 has had a Senior Reactor Operator's license and he has a 14 master's degree in health physics. Thank you. 3 r' !
'( ,N) 15 MR. BERGENDAHL: Thank you, Lew. Thank you, Ms.
16 Adensam and members of the NRC staff. I welcome the 17 opportunity to be here today. I think I speak on behalf of 18 everyone at the Perry Station when I say I am proud of what 19 we are doing at the plant, and we welcome the opportunity to 20 come here today and tell you about it. 21 First of all, I would like to make it clear that 22 First Energy has submitted a detailed, written response to 23 this petition. I will be highlighting a couple of the key 24 issues related to that written response. In addition, the 25 Perry Station is very similar in design to the River Bend j O ANN RILEY & ASSOCIATES, LTD.
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L 71
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- 1 Station, so much of the specific items that i:r. Bethay has
- O. 2i already discussed applies to our plant as well.
l- V l
, 3 But I am here to speak about the Perry plant 4' today. 'There's three topics I will be discussing. .First of 5 all, I am going to explain.how we are operating the Perry l- .6 plant. safely. Second of all, I am going to explain how we 7 are operating the Perry plant.as it was designed and 8 licensed.to be operated. And, third, I am going to discuss 9 the culture at Perry which. drives us to continucusly do 10 better.
11 My first graphic here illustrates these three 12 topics. Public health and safety is always the barrier we 13 will never cross. We will never cross thet boundary. I am 14 . going to discuss safety first, it is our number one 15 ' priority. Next, the design and licensing basis, that is the j 16 regulatory framework that keeps us with adequate margin 17 above and beyond the safety requirements. And, finally, 18 safety culture -- safety culture is the third topic I am I 19 ' going to discuss and what safety culture does is ensure that 20 we do even better. 21 First, safe operations. We know safe operations 22 .is our responsibility. Safe operations is my 23 responsibility. It is in our own best interest since our
'24 plant is located in our community. I live in Lake County.
2 5.. My children attend.a day camp down the street from the i ANN RILEY & ASSOCIATES, LTD.
.I Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
p 72 1: plant. 'Our plant manager,' his wife and seven children live (} ~2
'3 almost next door to this plant. We know safety is important. ;Our 800_co-workers are'the local community. We 4' have a vested interest in safe operations.
- 5. The way we keep our plant safe is we continuously
;6 ' monitor it,.very closely. A few months ago we detected'a 7 slight change in the xenon gas ratio in our cooling water.
- 8. .As a result., we systematically inserted each one of our 9 . control rods, one at a time, took hundreds of chemistry 10 samples, and we concluded we may have three pinholes in our 11, fuel' assemblies -- three rods may be effected. That's three 12 out of 50,000 fuel roads which make up our core. These 13 pinholes may not even be visible to the naked eye, but ou-14 chemical analysis detected a very slight change.
() 15 M
.y next graphic illustrates a little perspective 'i 16 or what we have seen with this gaseous activity. The first 17 bar1 illustrates the regulatory dose limits, and I have had 18 to cut this bar off because it is about 30 times higher than 19 what you see on this page here, but that is where the
-20 regulatory dose limits are.
21 The second bar is our technical specifications. 22 Our technical specifications are set such that we will l 23- remain at only 2 percent of the off-site dose limits, and 24 that technical specification is based on the assumption we 25 lose all of our waste treatment capability. If we had no
'O ANN RILEY & ASSOCIATES, LTD. '(_) Court Reporters 1025. Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
73 1 waste treatment capability, that activity at that level 2 ( ) would result in 2 percent of the regulatory limits. 3 We , because of our culture, keeping our community 4 safe, we set our own administrative limits, about 30 percent 5 of that technical specification limit, and our current value 6 today is about 3 percent of our own' administrative limit. 7 Negligible, very tiny changes in the activities in our
'8 plant. This situation has had no impact on our ability to 3 operate this Perry plant safely.
10 So what are we doing? We have addressed this 11 problem like we do any other situation in our plant, we use
-12 our Corrective Action program. Our Corrective Action 13 program ensures that we investigate, determine the cause and 14 fix any issue that arises in our plant. Part of my (m
( ,) 15 responsibility at the piant is our Corrective Action i 16 program. I am responsible for the Corrective Action ' 17- program, I am responsible for our Quality Assurance program, a l 18 and I am responsible for our Radiation Protection program. 19 We take this seriously. Our Corrective Action 20 program is a normal part of our business. This is a normal, ) 21 _ daily activity. We have an issue, we investigate it and we l 22 fix it. l 23 We are a large industrial facility. We have a lot
'24 of equipment, we monitor it every day, and if anything 25 changes, we investigate it and we correct it.
ANN RILEY & ASSOCIATES, LTD. [s -) Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
74 1 1 Now, continuing on this safe operations, you have l a (em) 2 heard a lot of discussions today about "what if" scenarios. l 1
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3 We do that, too. We constantly -- our staff is constantly l' 4 developing new questions, what if this, what if that. We
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5 analyze, we come to a conclusion and we move on. 1 6 I am not here today to talk about "what if" 7 scenarios. We are here to talk about the facts at the Perry 8 Station. The facts are there has been no increase in the 9 work area radiation levels in our plant -- no increase. And i 10 there has been no increase in the readings on our effluent 11 radiation monitors -- no increase. Those are the facts, we 12 are operating our plant safely. 13 Now, the second topic I said I would discuss is 14 our design and licensing basis. We are operating the Perry r~N (.vj 15 plant as it was designed and licensed to be operated. Our 16 plant is designed, like many -- like all other plants, to 17 meet hundreds of design specifications, ANSI Standards, ASME 18 Codes, hundreds. We met those codes, it is part of our 19 decign. 20 When you look at the overall design of our plaat, 21 what you will see is it is clear that operation with some 22 fuel defects was part of the assumptions. For example, the 23 walls in our plant are designed with shielding to 24 accommodate continuous operation with fuel defects. A waste 25 gas -- all our waste treatment systems are designed for ANN RILEY & ASSOCIATES, LTD. f'/) (_ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
75 l' .long-term operation with fuel defects, it is part of our [%d') ' 2 ' design- . Our radiation monitors and our procedures are all 3: set up'withithe assumption that there will be some fuel
'4 defect > source term.
5' Now, in addition to being designed for fuel 6 defects, our operating license basis explicitly refers.to 7 -fuel defects in our analysis. First_of all, if we' start l 8 with'the Nuclear Regulatory Standard Review Plan,'it clearly 9 points out that it is not possible to totally avoid all fuel 10 defects. Our own Safety Analysis Report, in the section, 11 _the category called Normal Operations, has a' limit on the 12 amount of fuel defects that we can have. That limit is 13 based on dose limits. It is specificallyLaddressed as part 14 of our normal operations that we will limit operations to a ( ,o '15 certain level of defects. 16 Our safety analysis report in Chapter 15 under the
.17 . discussion of transients, anticipated operational 18 occurrences that we may experience, specifically refers to 19 preexisting defects. As far as the potential impacts of the 20 hypothetical' accident, our. safety analysis report 1 addresses- .
21 fuel defects once again.
>22 The control rod drop scenario -- the assumption is 23 over.700, rods'are damaged in that scenario and the worst 24 case, loss of coolant scenario, assumes all of our fuel rods 25 'are damaged and even with all of our fuel rods are damaged ANN RILEY & ASSOCIATES, LTD.
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76 1 .ourDanalysis' shows ame will: still maintain : safety for our 2 -community. Worst case scenario, we are still bounded. (}
.3: I use this. graph here to summarize our. licensing
-4 and design basis. .Once~again public and health safety, the 5 . number'one barrier-we will never exceed. We are at normal
.fh operations today. Our normal operations licensing basis 7- assumes there"will be some fuel defects. If we were to move
=8 'from-normal operations 1through a transient condition, you 9, :are starting'with some fuel. defects. You move.from your-10 -normal operations to the scenarios, the transient scenarios
.11- and you. move to an accident: scenario the assumption is all 12 the fuel is damages and we are still within the public 13 health and safety. Clearly the barriers there,'the' analysis
'14 has been done.
f 15 Now in addition to our design and licensing basis, 16 they consider operations with fuel defects. Our technical 17 specifications provide additional assurance'that we protect 18 the community's health and safety. Our technical 19 specifications have specific operating limits which ensure
~20 ithe assumptions that went into our licensing basis remains 21 ' valid.
22 These technical specifications ensure that we have 23 limits and we use those limits, and we will shut the plant down if we reach those limits. Those limits are set so that 25 we will shut our plant down and we will shut it down before ANN RILEY & ASSOCIATES, LTD. C) Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
n 77 1 there is any impact to the community. That is the way the j 2 limits are set up. That is the way our technical 3 specifications are.
- 4. 'Our technical specification limits are set up such 5 that it'doesn't matter the cause, it doesn't matter how the 6 defect' started, it doesn't matter what transpired, if we 7 reach those levels we shut the plant down, period. Our 8 technical specification limit defines exactly how much 9 fission products due to fuel defects we-can operate with and 10 maintain safety in our community.
11 ~Now clearly with our design and licensing basis 12 specifically referring to these fuel defects, that is not 13 enough for us. The third topic I want'to talk about is our 14 culture. Our culture drives uc to do even better. This is
! 15' our community we are talking about here. .We are talking 16 about keeping our plant even safer.
17 All operators'are' trained and retrained at the 18 Purry plant to respond to a variety of different scenarios i 19 including changes in the fuel status. This is a group of 20- professionals that take their job seriously. Our operators 21 know about the awesome responsibility that they have. They 8 22' have a safety culture that is well embedded. As a matter of 23' fact, this past summer the Institute of Nuclear Power 24 Operations sent a 25-man team into our plant and they 25 praised our operator safety consciousness. It is built into i l l i
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V Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 J
i 78 I 1 our culture. 2 Now.how'do we measure safety? Now can our 3 operators impact it? My next graph explains one of the 4 -tools we have'. Our probabilistic safety analysis is a tool 5 out operators'use every day. This analysis allows our-n 6 operators to look at the impact of every piece of equipment 7 and' hot.to ensure we maximize safety. We have to change oil 8 on pumps. We have to grease breakers. Our operators look > a 9 at the impact of each of those components and schedule'the I I
.10- work such that safety is maintained to the maximum levels. I 11 NRC. expectations for probabilistic safety ,
1 12 assessment have been set. Our baseline plant design is l 13 better, exceeds the requirements, but because of our 14 cultur(, because our operators live in the community and (a): 15 were driven to do better in 1998 because of their attention 16 to detail on probabilistic safety analysis and the safety 17 impact of everything we do, we exceeded even our design 18 assumptions. 19 Now it is not only our operators. Our maintenance 20 craftsmen, our engineering professionals, the same safety 21 culture exists there. They use our probabilistic safety 22 assessment. They use the Nuclear Regulatory Commission 23 - maintenance rule. That's another tool. Our maintenance 24- . personnel, our engineering personnel closely monitor the 25 performance of .11 our systems that'are important to safety. [^' N ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
~
T- l l 79 1- This close analysis, this close monitoring, these high 2- standards have allowed us to achieve goals, make ( L 3' accomplishments in the' area of the NRC maintenance rule 4- which-is the'best in our region. It's probably the best in 1 5 the industry because of our culture, because our employees , 6 knowLthat it is their community. 7 Now a good measure of safety culture is radiation 8 o exp'sure reduction. I would like to' talk about that for a 9,. second. 'Our conservative actions, our conservative 10 operations and safety culture clearly shows through in our 11- ALARA performance -- ALARA, "As Low As Reasonably. 12 Achievable" -- is the philosophy we use in determining where 13 our radiation exposures can be and should be. 14 These tiny pinholes that we have detected in our () 151 16 fuel did not stop us from excellent performance in radiation exposure reduction. As a matter of fact, last year Perry 17 had its lowest radiation exposure ever. As a matter of
-18 fact, Perry's exposure last year was the lowest of any plant
'19 of its design ever.
20- My coworkers and I received an average of less 21 than 1 percent of'the acceptable limits on radiation
- 22 exposure. My next graphic illustrates'where we stand on j 23 radiation exposure reduction and ALARA.
24 'The regulatory limits are 5 rem:a year for nuclear 25- workers. We set our own limit -- one rem is our i
-[~h ANN RILEY & ASSOCIATES, LTD. \s- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
f 80 1 administrative control level. Last year, less than .1. Our I~' 2 employees -- we accumulated less radiation exposure than i I A 3 non-nuclear workers get from normal background radiation. ] 4 We also discussed here today the evaluation of the exposure 5 involved in going inside our containment right now, going i 6- inside our drywell, disassembling our reactor vessel and ! 7 searching out these three pinholes -- out of the 50,000 fuel 8 rods, searching through the fuel rods to find these three : 9 pinholes. The dose associated with going in just to do 10 that, compared to our current actions, which are to closely 11 monitor the situation, not expose our workers to any ! 12 unnecessary exposure, and disassemble the reactor next month 13 when it is scheduled to be disassembled, and not involve any 14 extra exposure. The ALARA analysis is clear. What we are () 15 .doing is achieving as low as reasonably achievable. 16 Now to conclude my remarks, this information I i 17 have provided should allow the NRC Staff to conclude that ! 18 this petition should be denied. , i 1 19 There is no health and safety. issue to be resolved 20 here. The public, our friends and families around the 21' plant, are at no risk from our current condition. There is 22 no risk because we operate our plant safely. We operate our 23 plant as it was designed and licensed to be operated and our 24 culture drives us to do even better. 25 Now again I would like to thank the NRC Staff for [T' 'sm ) ANN RILEY & ASSOCIATES, LTD. Cottrt Reportars 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
m 81
- the opportunity to be here to' day and all our listeners for
( F2 your attention,.and myself and the Perry staff will be glad 3 to answer any questions you may have or provide any 4 ~ assistance-we can provide in bringing this issue to closure. 5 'Thank you. 6 'MS. ADENSAM: thank-you, Mr. Pergendahl.
- 7. 'Does the NRC Staff have any questions-of 8- FirstEnergy?
9_ MR. CARUSO: I'have a couple of questions. 10 The first one was in its petition the UCS claimed 11 that movement of fuel rods could cause a power increase on i 12 the order of a factor of 10 in the local power peaking i 13 factor. Do you know if this is possible? Does anyone know 14 if=that is possible? () MR. BERGENDAHL: We believe it's clear from our 16 design licensing' basis, our operating licensing basis 17~ = analysis that_the assumption is we start failing fuel and j 18J that is our starting point and we enter the transients with ) 19 F i e 'lel already failed and our analysis shows regardless'of 20 ni and_how the defect or failure occurs, we are still'weIl- '{ 22 sitnin safety bounds, so there's a lot of different 22 scenarios of how or when the failure can occur. 23- MR. MYERS: I want to answer that one too. We 24 suppress the fuel, the defective fuel rod wherever we have 125 ' inserted a rod beside the defective fuel rods, so we have i ANN RILEY & ASSOCIATES, LTD.
.s Court Reporters 1025 Connecticut. Avenue, NW, Suite 1014 Washington, D.C. 20036 '
(202) 842-0034 1
i 82 11 suppressed those' rods already and we have administrative 2 . limits on the power changes now. f} ~
-3 We had a power change this weekend of less than 1 4- _ percent per hour so.we don't think that is -- it could be
~5. possible but it is.not very likely. Once again it is a 6- 'safetyLissue.
7L MR.'CARUSO: I think the question they're asking 8 is if you have a relocation of affuel rod because it failed, 9 could you cause a local power peak that was 10 times higher
'10 .than the power level before the. fuel rod relocated, and I 11 would ask Entergy do you know if that is possible also. '
12 .MR. MYERS: Paul would know. 13 MR. BORDLEY: Yes. I am Paul Bordley. I am the 14 reactor engineer at the Perry plant.
.P 15 I really think what you are going after is if you-16 fully withdraw a control rod, can a local power change by a 17 factor of 10. A factor of 10 seems a bit high.
18 You can obviously increase the local power, say, 19 ~from 4 kilowatts a' foot to 7 or 8 kilowatts a foot with a 20 . rod withdrawal , yes. That occurs during normal operation. 21 I don't know if you have a further question or
.22- concern.
23 Now with the -- as Mr. Myers said, when -- with 24 our current three projected defects suppress any power 25- changes are obviously, that could occur are obviously fT d ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
l 83 1 ' reduced. 1 f~'N 2 MR. CARUSO: Okay. I had one question for Harry. l
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f 3 You are going into your outage and you are going to try to 1 4 identify the three rods that you think are the pinhole leaks j 1 5 and if you go into the outage and you remove the bundle that you think the pinhole leaks are located in and you cannot 7 identify the exact pin, what will you'do? 1 8 MR. BERGENDAHL: We will replace that fuel l 9 assembly. I 10 MR. CARUSO: The whole fuel assembly?
)
)
11 MR. BERGENDAHL: That's correct. { 12 MR. MYERS: Yes, that is correct. 13 MR. CARUSO: How certain are you that you will be 14 able to identify the correct fuel assembly? rs ( ) '15 MR. BORDLEY: Going into the outage we are going 16 to conduct a fual sipping campaign for all exposed reload 17 fuel. We are using one of the current state-of-the-art the)- 18 call it fuel sipping system to find the defect, and what we 19 will do is when we defect a fuel defect with the sipping 20 equipment, which is being run by a subcontractor, we will 21 take the fuel assembly down to our fuel handling building, 22 dechannel it, and then we will perform a post-irradiation 23 exam poolside of the fuel assembly and in that 24 non-destructive examination that occurs you send the fuel 25 rod -- first you do a visual of the outside of the bundle. -I ANN RILEY & ASSOCIATES, LTD. (_-) Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
i I i 84 1- You take a look at the rods, you know, see'if there is an () 2 3, obvious defect there. if not, you take rods within the bundle and you put them through an eddy current in an 4' ultrasonic tester, looking_for the defect site. The goal is 5 to find the' defect so that we can take a visual look at it 6 and most.of the time a visual look at the defect will tell 7 us theLcause. 8 MR. CARUSO: Okay, so you are not going to just 9 depend upon radiological sipping to identify the leaking 10 fuel? 11 MR. BORDLEY: No. No, we intend to back that up 12 with-post-irradiation exam in our fuel building. 13 MR. CARUSO: Okay. One more question for both 14 Perry and River Bend. Do you have any estimate of the A
- (
) 15 burn-up of the fuel that is leaking? Do you know -- do you l 16 have an idea of where the fuel is located? Do you have an 17 idea what the fuel burnups are of those particular elements
- 18. or pins?
19 MR. BORDLEY: Okay -- well, at Perry to date, to , 20 determine the age or the exposure of the leaking assembly 21 you need to see certain isotopes and the defects at Perry 22 have not released those isotopes so ours are currently very 23 tight, so the absence of evidence confirms that the defects 24 remain extremely tight. 25 MR. CARUSO: No, I guess what I's asking is you I (O
's /
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85
- 1 ,think you'know which elements have the defects --
V[ j- 2 MR. MYERS: No . - We can't. We don't. We know the 3 cell. We won't know where the defect's at until we go look.
~4 MR. CARUSO: Okay, so you don't have any idea of 5 Twhether these are once, twice, three times burned fuel?
6 MR. MYERS: The answer to_that question is really
.7 no.
8 MR. BORDLEY: Within a control cell we'll have -- 9 it's not all one' type' fuel assemblies. In two of the 10 control cells which are prime suspect locations we have fuel 11 that is in its first cycle and its third cycle. In the 12 ccher one we have fuel that is in its first cycle and its 13 second cycle and without the isotopic'information, we can't 14 define it any further at this point.
- f i 15 MR. CARUSO
- Okay. what about River Bend?
16' MR. KING: At River Bend we, as Perry, have done 17 'some testing to identify the cell locations. As far as the 18 specific pin, until we do the actual 100 percent core 19_ sipping and the investigation analysis won't be able to get 20 to that level of detail. However,'we have through some 21 chemical sampling identified that we believe the majority of 22 ours are in-the first burn for the cells that have been 123 : identified. 24- MR. CARUSO: The new fuel? 25 MR. KING: Yes. .O' ANN RILEY & ASSOCIATES, LTD. ds/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
r 86 1 MR. CARUSO: Let me ask, do you think that you
~
[ 't 2 might extend your PIE examination beyond the cells where ycn () 3 think there is a problem to look to see if there might be 4 any incipient problems in other fuel elements of the same 5 batch or a similar batch? 6 MR. KING: Yes. We plan to make sure that our 7 investigation and cause analysis takes that to adjacent cell 8 locations and do a full investigation. 9 MR. DONOGHUE: One more question for both 10 licensees. Will the examination of fuel be done before the , 11 next startup? 12 MR. BETHAY: Yes, for River Bend clearly it would 13 be done during the shutdown and then our outage is starting 14 in April so these examinations will be completed prior to ;
,\
() 15 startup. 16 MR. DONOGHUE: So we will know what is going on as i i 17 far as what kind of fuel is involved and what a defect is. 18 MR. BETHAY. Well, we will certainly have a lot 19 more information than we do now. 20 MR. DONOGHUE: Obviously -- 21 MR. BETHAY: So hopefully whether we'd have a 22 definitive, iron-clad this-is-what-caused-it, we might not 23 have that prior to startup but we certainly, if it debris, 24 if it's a manufacturing problem, a whole lot better idea if 25 not conclusively. [' ^) ANN RILEY & ASSOCIATES, LTD. C/ C:trt Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
87 1 MR. BERGENDAHL: That's correct for Perry too. ( ,) 2 We'll be analyzing fuel as it is removed. x_/ 3 MR. DONOGHUE: Thank you. 4 MS. ADENSAM: I have a question for Perry, if I 5 may. I understood from Mr. Bethay at River Bend that they 6 had estimated something on the order of nine person-rem to 7 remove the leaking fuel. In your estimates in looking at i 8 your ALARA concerns, did you come up with a similar estimate { 9 and was it something that would be simply more associated i 10 with opening up and moving fuel regardless of whether or not l 11 it were leaking or whether it's associated with leaking ) 12 fuel? 13 MR. BERGENDAHL: Our response is similar to River 14 Bend's. We estimated on the order'of about 15 rem to /~ j ( ,Y) 15 dissemble and retrieve the defective fuel rods and that ! 16 would be in addition to our refueling outage dose, so it 17 would be extra dose and would not be as low as reasonably 18 achievable. 19 MS. ADENSAM: Extra dose in the sense that you 20 have opened the vessel twice or because of the damaged fuel? 21 MR. BERGENDAHL: Twice, because to remove the fuel 22 now we would still need to do our refueling outage where we 23 would have exposure to our workers twice as opposed to once. 24 MS. ADENSAM: Okay. 25 MR. MYERS: I'm sorry, go ahead, Elinor -- ANN RILEY & ASSOCIATES, LTD. [)/ x- ' Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
r-88 ) l' MR. BERGENDAHL: However, E want to re-emphasize-(( ~2l the ALARA-philosophy:we are constantly evaluating the 3 impacts, and if something were'to change, we would still f i 4 shut the plant down, if that was the right thing to-do,'and l
~.5 our exposures were-being impacted.
6 MR. MYERS: Our next refuelling. outage we are 7 ' working on these effective fuel pins. That's one of our
~8 most: aggressive does goals ever, and we haven't changed l 9 .that,.so we' don't anticipate any increases in radiation.
'10- MS. ADENSAM: Thank you.
11 If there are no.further questions, I think we can 12 proceed to our public comments. When we spoke to our 13 colleagues at Baton Rouge earlier, they said there was no 14 one there who had an interest in making a comment. I'd like ( - 15 to find out if that's still the case. Mr. Marshall? 16 MR. MARSHALL: . We have no participants that would 17 like to speak at this time.
.18 MS. ADENSAM: Okay. Thank.you very much. )
19 We'd like to go now to Lake Erie. I understood ) 20 there was at least one individual who had indicated an
'21 interest.in making a comment. Mr. Clark, could you let us 22 know if there's anyone else?
23 MR. CLARK: There's two at this time. 24 Can you hear me clearly, first? 25' MS. ADENSAM: Could you speak a little more into ANN RILEY & ASSOCIATES, LTD. O- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036
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l t 89
'l the mike, Mr. Clark?.
[0
Y 2 3 basically.
MR '. CLARK: I'm going to have to hold the mike
.Can you hear me? J 4 MS ADENSAM: Yes. That's very good. Thank you.
5 And if_you could ask the participants to give us their names as clearly as possible, and also'to hold the mike if they 7 'would, please. 8 MR. CLARK: The first one to speak I believe was i 9 Connie Klein. l 10 MS. KLEIN- I' don't have any comments, but I have R11 several questions. 12' MS. ADENSAM: . Excuse me. Could you state your 113 name, please? 14 MS. KLEIN: Can you hear me? () ~15 MS. ADENSAM: Could you state your name, please? 16 MS. KLEIN: Connie Klein. j 17 MS. ADENSAM: Thank you. 18 MS, KLEIN: Connie Klein. Connie Klein.
'19' MS. ADENSAM: Thank you.
20 MS. KLEIN: Okay. My'first question is, isn't it 21' true that failed: fuel or defective fuel cannot be dry 22 . casked, and also cannot -- could not.be shipped to a
- 23. repository if one existed?
24 MS. ADENSAM: Mrs. Klein, that's a good question. 25' We are not prepared at this time to. respond. However, our
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d 90 1 expectation is that;we will be responding to you through-() ~ 2. 3
.other forums. But go right ahead with your questions.
That's perfectly all right. I 4 MS. KLEIN: In light of what the representatives 5- from FirstEnergy.said, can we assume that Perry reached its 6 tech. specs in November or leading up to November of 1993 7- zwhen the' reactor was shut down because of' seriously 8 defective fuel, apparently more seriously defective than
.9 what's being assumed now?
10 Do you understand-my question? Were the tech ! 1 11 specs exceeded, and is that the reason -- there definitely 12 was worker exposure during that period. )
- 13. MR. HEGRAT: I can answer that question. My name 14 is Henry Hegrat. I'm a regulatory affairs manager at Perry. !
() .15 ' After our forced refueling outage, we did not 16 exceed tech' spec limitations for the fuel failure that we 17
~
experienced at that time, and we did-replace that fuel 18' failure at a regularly scheduled refueling outage. 19' MS. KLEIN: You had an unscheduled outage then. l
'20 MR. BORDLEY: Right, in January 1993 we did take l 21' an_ outage. We were still.at a fraction of the technical 22' specification limits. Plant management at that time made 23 the decision to.take the outage to remove the fuel defect.
- 24. The intention at that time was to minimize contamination and 25 dose in the plant for the future.
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na 91 1 MR. MYERS: Ms. Connie, once again I think it's ( v}. 2 real important to note that, you know, we have
-3 _ administrative limits that's well belcw our regulatory 4 limits, and, as I. stated earlier, Perry is very committed to 5 keeping a very low dose in our plant and to keeping the 6- plant in good material condition, so we react promptly to 7 those' situations.
8 Thank you for that comment. 9 MS. KLEIN: Okay. Just one more question. This 10 is directed to the NRC. When you're dealing with.obviously
'11 really serious fuel-defect situations like those at 12 Palisades and Point Beach, I know you're not going to answer 13 me, but I just want to get this on-the record, is it the 14 NRC's contention that when fuel pellets are exposed, when a s
() 15 16 situation like fuel drop, is it the NRC's contention that the reactor should not be prophylactically shut down under 17 those kinds of really serious fuel-defect situations? 18 MR. WERMEIL: This is Jerry Wermeil. I'm the new 19 chief of the Reactor Systems Branch. Let me try to answer 20 your question within a context once again of the NRC's 21 regulations and.the license that the plant is operating 22 under. 23 -If a fuel-damage situation or a fuel-defect 24 situation results in exceedence of any of the requirements 25 either within our regulations or within the plant's license
-[~')'
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ANN RILEY & ASSOCIATES, LTD. Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
1 i 92 1 'itself, then the plant would have to act accordingly. If_ (J :2 the license was in some.way compromised or a technical 3 specification could not be satisfied, the licensee addresses 4 'itLin accordance with the conditions specified in the j 5- license. If the'NRC regulation is somehow impacted or L6L somehow affected by the situation that arises, once again 7- the'NRC'and the licensee address the situation and would 8 react accordingly. 9' The_ specifics of Palisades, and I believe you 10 mentioned another plant I am not aware of, but it strikes me 11 that if a plant has experienced say a fuel-rod drop or some 12 other accident that has exposed the fuel, it will be forced 13 to continue its operation only within the confines of its 14 license and our_ regulations. ( 15 MR. CLARK: Thank you. Who's next? 16' MR. BIMBER: I am Russell Bimber. I worked as a 17 - chemist locally for about 40 years, and have worked a little q 18 .bitHas a volunteer connected with radiation safety around 19 the Perry plant. 20 Mr. Lochbaum seems to' assume that operation with 21 pinholes in the_ fuel is not permitted, and yet there was an 22 NRC. letter handed out at the beginning of this hearing which 23- . appears _to-indicate this was previously considered. Now 24 _there seems to be some discrepancy between-Mr. Lochbaum 25 citing.NRC documents'which appear to cover manned operation ANN RILEY & ASSOCIATES, LTD. O- Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
93 1 .with defective fuel and the statements from both the reactor () 2 3 operators and the NRC that operating with defective fuel has previously been considered. Now I'm puzzled by this
- 4. Jiscrepancy which led to this hearing. Would you care to 5 comment? I 6 .MR. WERMEIL: This is Jerry Wermeil. Let me try 7 to put a little context on what I think the discrepancy is.
8 Fuel failures or some amount of leakage from the fuel is 9 assumed as part of the license basis for every nuclear io powerplant. That was clearly pointed out, and I believe it j 11 was clearly agreed to by Mr. Lochbaum in his statement when 12 he indicated that there are technical specifications under i 13 which any nuclear powerplant operates that limit the amount. 14 of leakage and the amount of activity in the reactor coolant (j ( g 15 that's permitted for normal operation. 16 What I believe Mr. Lochbaum is trying to say is ; 17 .despite the assumption on failed fuel that the-NRC staff and 18 perhaps the licensees have not effectively taken the 19 consideration of existing fuel failures or leakage within 20 the fuel consistently in their accident analysis 21 assumptions. In other words, when the licensee analyzes the 22 set of accident analyses and transient conditions that the 23 NRC. imposes on a licensee as part of its design basis, this 24 assumption that there is leakage or that there are some fuel 25 failures hasn't been properly accounted for in those ; 9 l l
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1 i 94 i 1 snalyses. And that I believe is the issue that.we are going lV 2 t'o address and that the licensees have addressed in their L) 3 response to the petitions. 4 MS. ADENSAM: Mr. Clark, were there any other
'5 comments from the participan<, in Lake Erie? !
I 6 MR. CLhKK: Yes. Yes, we have another one. Yes, l l 7- we do have another at this time. 8 MS. ADENSAM: Okay. i 9 MR. CLARK: Go ahead and state your name. l l 10 MR. FETERAL: Yes. This is Jack Feteral. I live l 11' on Clark Road just down from the plant. I have two j 12 questions. The first question is about the materials and 13 the constructicn of the plant, and the second is about 14 public safety. I will give you both questions first, and () 15 then wait for your answer. 16 The first question, can th'e leaking radioactive I 17 material in the core degrade or cause increased electrolysis j 18 of any of the plumbing and seals in the loop? Can the 19 released radioactivity in the loop degrade the alloy or 20- material or change the atomic structure of the material in 21 the loop, and if so, how much? And can this cause 22 down-the-road problems? 23 The second question, about public safet y . There 24 was a road connecting the western part of Perry Township 25 with the eastern part of Perry Township that was cut by the h'
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p , l 95 l 1 construction of the plant. We were verbally promised at one 2 point that it would be reconstructed or there would be a 3 road put in around the plant.
'4 If there was an outside leak of radiation 3 1
5 transported by.a north-northeast wind and a wind shear 6 occurred at Route 20 and Center Road from the southwest, 7 residents would not be able to'use the closed evacuation 8 escape route. This construction or reconstruction of this L 9 road would also make movement of school buses and residents 10 between Perry and Madison much safer. Are you going to do 11 anything about this in the future? 12 Thank you. 13~ MS. ADENSAM: Mr. Feteral, we weren't really 14 prepared, quite frankly, to respond to questionsoof that j
/~5 !
( )v 15 nature, although they're very good questions. We will try ' 16 and respond to those. If you could let Mr. Clark know what i 17 your address is, we can try and get back to you, or pick i 18 another forum to respond to those questions. 19 MR. MYERS: I would like to respond to the first. i 20 MS. ADENSAM: Okay, Mr. Myers. , 21 MR. MYERS: 'As we stated earlier, we believe in 22 maintaining the plant in very, very good material condition, 23_ and it is in good material condition. One of the things we 24- , in every refueling outage, we do what we call i~.pections 25 of all of our systems. So if we were to see any degradation
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1 96 1 1 of our systems, we're constantly _ monitoring and looking for (~'O 2 that degradation. We do that every outage. Once again, (,./ 3 that's.the safety culture we're talking about. So we would-4 see that if that were to happen. Okay? j 5 MS. ADENSAM: Thank you. l 6 MR.-CLARK: Are there any other comments? {
~7 That concludes the comments from the Lake Erie l
8 College site. / 9 MS. ADENSAM: Thank you, Mr. Clark, and we'd like 10 to thank the individuals who have participated this 11 afternoon I I 12 I would like to go to Cleveland, to Ms. Lipa. 13 MS. LIPA: Okay. We're here. Can you hear us?
]
14 This is the Cleveland site. Can you hear us? l ir) 15' MS. ADENSAM: Yes, we can. Thank you. j 16 MS. LIPA: This is Christine Lipa. Good. We have 17 a person here who would like to speak. Hold on. 18- MR. ELLISON: Hi. My name is David Ellison. I'm 19 a registered architect, and I practice here in the city of 20 Cleveland. 21 I've been around nuclear scientists all my life. 22 My father was a scientist with Sandia Laboratories, and I've 23- been talked down to by nuclear industry representatives and 24 regulators almost all my life. So I would like to make a 25 _few comments here, and I really don't expect any meaningful -; ANN RILEY & ASSOCIATES, LTD.
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97
~1 answers from you.
( 2 Similar to the-comment from Lake Erie College 3 regarding the discrepancy between the' accident scenarios and 4 what -- which are planned and the ones that might actually 5 'take place as a result of these leaking fuel rods, I'm 6 concerned that neither the plant operators or the NRC 7 acknowledge that there'is this discrepancy'whether or not 8 the occurrence might be dangerous or represent a threat to 9 public health and safety, as Mr. Lochbaum suggests. So the 10 fact that this discrepancy has not been adequately addressed i
- 11 in either of the comments from the power companies or the 12 NRC is a problem for me. And I would hope that in the~
13 subsequent response to this petition that this matter be
. 14- explained a whole lot more'in detail. -j
) 15- The next comment I have is for the NRC, and hera-16 .in Ohio we have a history of the NRC enforcing the 17 regulations on many licensees, in fact, hundreds of 18 licensees here, and many of those licensees have operated 19 within the confines of the regulations on the radioactive j 20 materials that they're dealing with. And despite the fact z 21 that the NRC has.been regulating them or the Atomic Energy 1 22 Commission before the NRC, that has resulted in 23 contamination of Ohio's environment that was totally 24 unexpected. For instance, our sewer plant here is totally 25 contaminated by cobalt which was released within ehe '
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.r : 98 1 regulations allowed by the. regulators, but which fx c i 2 reaccumulated.
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3 And so my concern in this case is that the 4 contamination that results from the radiation released from 5 leaking fuel rods would present a problem for us later on 6 when FirstEnergy or whatever company owns the powerplant, by 7 the time it gets to the end of its license, walks away from 8 the plant and leaves it in the lap of the public around 9 here, and what greater exposure and what greater expense 10 we'll have to incur at that time as a result of the 11 inattention to leaking fuel rods at this time. 12 I would also like to point out that the dose 13 limits that people are allowed to receive vary widely from ; 14 agency to agency, and that to have charts that indicate that (3 ( ,) 15 you're well within limits of dose exposure while not-16 pointing out that the dose exposure for a person working in 17 a powerplant might be totally different than a person 18 working to decommission and get rid of that powerplant are 19 two different things, and that it's not a meaningful thing j 20 and it's misleading, and I don't really appreciate it. 21 The other issue that occurred to me while this { 22 hearing was taking place is why the manufacturers of these 23 rods aren't here and why there isn't more historical ! 24 information on what these rods do and why automatically we ! 25 don't have an answer for Miss Klein's question regarding dry f~')
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99 1 casking and the transportation or long-term storage of /m/ 1 2 leaking fuel rods. \_) 3 So when the fellow spoke about the culture of the 4 powerplant there at Perry, from my point of view, the 5 company that originally built that plant did something very 6 sneaky when they put the shovel in the ground the night 7 before regulations in Ohio went into effect which would have 8 caused an environmental review of the siting of that 9 powerplant and supported public regulation that would have 10 prevented that powerplant from being built, and then they 11 proceeded to carry on business in a way which led them to 12 near bankruptcy. 13 And which has resulted in the Perry Power Plant 14 being listed on a list of lemon power plants in the country, r\ i j 15 and which continues to be on the list of power plants which 16 are being watched by the NRC as some of the least well 17 managed in the country. And this, to me, despite the fact 18 that this fellow tells us that there's this culture of 19 safety at the plant, and that they are really looking out 20 for us, despite that, despite how many times we have been 21 told how much care there iL taken there, my perception, and 22 I think that the large percentage of the general public's 23 perception of power -- nuclear power plant safety, is well 24 described by the Stinsons and the Stinsons' role at the 25 power plant there in Springfield. So when you talk about ( ANN RILEY & ASSOCIATES, LTD. \_/ Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034
100 1- nuclear _ power plant culture, this is what I kind of feel l[ V 12 like. 3- I thank you for listening to my. comments, and like 4 I said before, I don't really expect to have any answers 5 from this, but I do fully expect for the NRC to continue
'6 accommodating whatever the power generators need when it 7 comes time to help them operate their plants, and I do fully 8~ expect somehow to end up as a member of the public' helping 9 them pay for the bailout of whatever company follows 10 FirstEnergy and the rest of-these power companies _that walk 11 away from their nuclear power plants when it comes time to, 12 you know, entomb them or decommission them or1whatever we 13 are going to end up doing with them, wtan they have either 14 melted down or finished their useful -- well, that would be p)
(_,.. 15 questionable -- but' finished their profitable, for the 16 companies. So thank you. 17 MS. ADENSAM: Thank you for your comments, Mr. 18 Ellison. 19 Ms. Lipa, is there anyone else? 20 MS. LIPA: No , that's it from here. 21 MS. ADENSAM: Oh. Okay. Thank you. I. guess -- l I 22 MS. LIPA: There's nobody else.. 23 MS. ADENSAM: Thank you. 24 I' guess I would like to ask if anyone, any public
^ . 25 participants here in headquarters in Rockville would like to I\ ANN RILEY & ASSOCIATES, LTD.
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p 101 1 1- make some comments. I believe I understood Mr.~ Butler ~ l f[~)
%J 2' wanted to make some comments? '
3 MR. BUTLER: Thank you for'this opportunity to 4 address the panel. My name is John Butler, and I am senior 5 project-manager.with the Nuclear. Energy Institute in 6 Washington, D.C. 7 Let me-start by stating my support for the actions 8 taken so far by the NRC in responding to the 2.206 9' petitions. NRC's preliminary evaluations have determined 10 that the issues raised in the petitions do not warrant or do i 11 not involve an urgent safety issue and do not warrant the 12 actions requested by the petitions. i 13 We understand that'this informal hearing was i 14 granted to allow UCS an opportunity to fully air their
) 15 concerns and to allow'the NRC to make a fully informed. {
16' decision on the issues raised in the petitions. The 17 presentations by Energy and FirstEnergy have confirmed that l 18- the Perry and River Bend Plants have and continue to operate 19 in a manner that is fully within their licensing basis and
- 20. within the applicable NRC regulations.
21 The presentations have also reminded us that the 22 licensing bases for these plants are founded on a principle 23 of multiple levels of conservatism and bounding analyses 24 ~ that are part of the defense-in-depth philosophy. 25 The UCS noted that we are discussing one of three gh ANN RILEY & ASSOCIATES, LTD.
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'l ~
102 1 barriers, the fuel, the- other two being the RCS piping and [)- 2 . containment. I think the presentation did a good job in 3- pointing out,that minor fuel defects do not threaten or 4 compromise worker safety or threaten public health and
'5 safety.
6 I think the presentations also noted that -- did a 7 good job showing that the current regulatory structure 8 protects public health and safety. This structure includes 9 detailed analyses that cover a range of actions from normal l 10 operation, transients and accidents that include fuel 11 problems ranging from minor fuel defects, such as Perry and 12 River Bend.have been experiencing, to complete fuel failure. 11 3 These analyses have been studied extensively by [ 14 the licensees and the fuel vendors, and have been studied ( 15 and reviewed extensively by the NRC. There has been no new 16 information introduced as part of this hearing that would i 1 17 call into question the licensing bases of these plants or 18 any commercial nuclear power plants. 19 So, in closing, I urge the Commission to come to a 20 speedy resolution in issuance of a director's decision that 21 denies in full the UCS petitions. 22 Thank you. 23 MS. ADENSAM: Thank you, Mr. Butler. 24 Is there anyone else? 25 I would like to make one comment in response to [~
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I i 103 1 the individuals at our. remote sites who are expressing () 2 3 concern about the discrepancy'between Mr. Lochbaum's assessment and that presented by the licensees, and this is 4 the'very point of the. Staff's evaluation and why we are l 5- trying to resolve that discrepancy, and we would hope that 6 the_ director's decision will accomplish that objective. 7 Since'Mr. Lochbaum has not been able to stay with' l 8 us , I guess we could_go on and ask if Energy has any closing ! 9 remarks it wishes to make. ! l 10 MR. TITUS: I just have a few brief comments in l l 11 closing. 12 We believe the current NRC rules and regulations 13 provide a firm basis to ensure that nuclear power stations 14 are operated safely, with defense-in-depth to protect the-(A )- 15 health and safety of the public, and the River Bend Station 16 is'in compliance with these rules and regulations. 17 In addition, we have provided in-depth technical 18- ~information that we believe demonstrates that the UCS 19 petition has no merit. In fact, operation of nuclear power 20 plants with minor fuel clad defects is not a new issue at 21 all, and the UCS has not in our view put forward any new 22 substantive information regarding this subject. Therefore, 23 we believe the UCS petition should be denied.
- 24. While we believe these points are compelling, I 25 want to assure you that our commitment to safety goes well
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7 104 1 beyond the rules and regulations. St. Francisville,
.; T 2 Louisiana is a very small community, and a significant
.w)
'3 number of the 770 River' Bend employees live in and around 4 the local plant area. Our employees are heavily involved in 5 the local community, from partnering with the local high 6 school to participating on the community development 7 foundation and being active leaders in the United Way, and I 8 could.go on and on and on. Our employees understand the 9 importance of the public trust they must uphold to run River 10 Bend safely, because it bears directly on protecting the 11 health and safety of their friends and neighbors and, l 12 indeed,_their very own families.
)
-13 Thank you.
]
- 14 MS. ADENSAM: Thank you, Mr. Titus.
n 15 () I'd like to go to Mr. Myers of FirstEnergy. j 16 MR. MYERS: I'd like to start out by thanking the 17 Nuclear Regulatory Commission for this opportunity to speak.
- 18. MS. ADENSAM: Thank you.
19 MR. MYERS: I would particularly like to tell you 20 about our perspective on the fuel situation. I trust our l 21 discussion today has provided you some useful information on 22 our performance. As stated earlier, the defective fuel rods 23 will be replaced in our March regularly scheduled refueling 24' outage. The best course of action is continue to monitor 25, the situation, and most importantly, to manage it t ANN RILEY & ASSOCIATES, LTD.
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105
'l successfully.
LI'T 2. As I stated, if.there is need, we will shut the X_./ 3 plant down earlier. After reviewing all the technical 4 information today, one thing remains clear: The Perry Plant 5' continues to operate safely, within our design basis, within 6 our. licensing limits, and we are having no impact on our ! 7 employees or to the community, and-I am personally committed i 8 to ensuring that it stays this way. 9 Thank you. ; 10 MS. ADENSAM: Thank you. l 11 I would like particularly to thank our public I
~ 12' participants for their comments,.and I apologize for our not 13 being -- having the right people here to answer all of their 14 questions. We will try and provide those responses in the
() 15 appropriate forum. 16 If there are no other comments from anyone, I 17 think this closes our hearing. .Thank you. 18 MR. MYERS: Thank you. 19 [Whereupon, at 3:40 p.m., the hearing was 20 concluded.] 21 22 23 24 25 i [ . ANN RILEY & ASSOCIATES, LTD. U' Court Reporters 1025 Connecticut Avenue, NW, Suite 1014 Washington, D.C. 20036 (202) 842-0034 c_
REPORTER'S CERTIFICATE ThisLis to certify that the attached proceedings before che United States Nuclear Regulatory Commission in [%s) the matter of: 1 ME OF PROCEEDING: INFORMAL PUBLIC HEARING l ON 10 CFR 2.206 PETITION j
-l l
I PLACE OF PROCEEDING: Rockville, MD were held as herein appears, and that this is the original () transcript thereof for the file of the United States Nuclear Regulatory Commission taken by me and thereafter reduced to typewriting by me or under the direction of the court reporting company, and that the transcript is a true and accurate record of the foregoing proceedings. sof S / Mark Mahoney r Official Reporter Ann Riley & Associates, Ltd.
e . [ \ UNITED STATES [ a, NUCLEAR REGU!.ATORY COMMISSION 5 ! WASHINGTON, D.C. 3008HelH O \...../ October 29, 1998 Mr. David A. Lochbaum Union of Concemed Scientists 1616 P Street NW, Suite 310 Washington, DC - 20036-1495
Dear Mr. Lochbaum:
I have received your Petition that you submitted on behalf of the Union of Concerned Scientists (UCS), dated September 25,1998, addressed to L. Joseph Callan, Executive Director for Operations, U.S. Nuclear Regulatory Commission (NRC). The Petition requests enforcement action to require an immediate shutdown of the River Bend Station (RBS) and that the facility remain shut down until all. failed fuel assemblies are removed from the reactor core. The RBS licensee, Entergy Operations, Inc., had recently filed NRC Daily Event Report No. 34815, in which it reported "a possible defect in fuel cladding." As an alternate action, you also stated that RBS could be restarted following the proposed shutdown after its design and licensing bases were updated to permit operation with failed fuel assemblies. Additionally, the Petition requested a public hearing to present new plant-specific information regarding the operation of RBS, as well as to discuss a UCS report dated April 2,1998, entitled " Potential Nuclear Safety Hazard / Reactor Operation Wdh Failed Fuel Cladding." .O . As the basis for your request, examples were cited in the Petition (summarized below) where, in your opinion, the RBS Updated Safety Analysis Report (USAR) does not allow for operation with pre-existing fuel failures: (1) Integrity of the fuel barrier is an explicit criterion in addition to radiation requirements, and RBS is violating "the spirit, if not the letter, of [USAR Section 15A, Table 15A.2-4) Criterion 4-2 since the fuel barrier has already failed, albeit to a limited extent." (2) The USAR description for six design-bases events includes either the statement that the fuel barrier maintains its " integrity and functions as designed," or that "no radioactive material is released from the fuel," as a consequence of the event. It is your view that the analyses associated with these events " appear {s) valid only when the River Bend Station is operated with no failed fuel assemblies." Your Petition further reasserted the UCS position that nuclear power plants operating with fuel cladding failures were potentially unsafe and were in violation of Federal regulations. In its April 1998 report, the UCS stated that it has not been demonstrated that the effects from design-bases transients and accidents (i.e., hydrodynamic loads, fuel enthalpy changes, etc.) prevent pre-existing fuel failures from propagating. Therefore, you concluded that it was possible that "significantly more radioactive material will be released to the reactor coolant system during a transient or accident than that experienced during steady state operation." In
4 David A. Lochbaum addition, you also stated that, by operating with possible failed fuel cladding, RBS is violating its licensing basis for the radiation worker protection (as low as reasonably achievable [ALARA]) program as it is described in USAR Sections 12.1.1, " Policy Considerations," and 12.1.2.1,
" General Design Considerations for ALARA Exposures."
When the staff received your Petition, it conducted a preliminary evaluation to determine if an urgent safety issue was involved that warranted the requested action. Although you raised important concems in your Petition, the staff has concluded that the Petition uncovered no urgent safety problems that warranted immediate action by the NRC. Technical Specifications (TS) limits on reactor coolant system (RCS) activity typically account for a small fraction of failed fuel, which can be expected during normal operations. These limits are set to values of RCS specific activity, which assure that the radiological consequences of postulated design-basis accidents are within the appropriate dose acceptance criteria. At RBS, operation with a minimal amount of fuel cladding damage is allowed, provided the licensee continues to meet RCS chemistry requirements of TS Section 3.4.8. Furthermore, the Petition did not include any information indicating that RBS has operated outside its TS limits. Consequently, your request for enforcement action to require the immediate shutdown of RBS is denied. The licensee has taken actions to address the suspected condition of the fuel assembly, including the insertion of control rods to " isolate" the fuel assembly in order to minimize reactor coolant activity levels. The NRC has been closely monitoring events at RBS and will continue to ensure that there is no undue risk to public health and safety. In your Petition, you also requested that the NRC conduct an informal public hearing in order to "present new information on reactor operation with failed fuel assemblies" as a follow up to the April 1998 UCS report, as well as to provide pir.nt-specific information regarding the operation of RBS. In order to ensure that potentialissues relating to this material are appropriately addressed, the NRC is hereby ofrering you an opportunity to present the new information referred to in your Petition at an informal public hearing. The public hearing will also allow the j licensee and public to present other pertinent information, as well as provide a means to solicit ; questions from all participants. The information gained from the hearing will subsequently be ! used by the NRC in evaluating the issues raised in your Petition and the eventual rendering of a Director's Decision pursuant to 10 CFR 2.206, ! I in order to assist in facilitating discussion on the relevant issues during the proposed informal ! public hearing, the NRC is requesting that you provide, in advance of the hearing, the new l information on reactor operation with failed fuel assernblies you intend to present. This will ! allow meeting participants to review the material and better prepare for the hearing. l If you wish to accept this offer, please contact Mr. Robert J. Fretz, the NRC Petition Manager, at (301) 415-1324 in order to establish a mutually agreeable date, time and location for the informal public hearing. Mr. Fretz will also serve as our point of contact for the information we have requested in advance of the hearing. O ; I
)
j j l David A. Lochbaum O l I i Your Petition has been referred to me pursuant to 10 CFR 2.206 of the Commission's I regulations. As provided by Section 2.206, action will be taken on your request within a reasonable time. I have enclosed for your information a copy cithe notice that is being filed
-{
I with the Office of the Federal Register for publication. I have also enclosed for your information a pamphlet on the public petition process. Sincerely, i Area- Q. h Frank J. Mirahia, et irector Office of Nuclear Reactor Regulation
Enclosures:
- 1. FederalRegisterNotice
- 2. Pamphlet on the Petition Process cc w/ encl: See next page o .
D r l
U Entergy Operations, Inc. River Bend Station cc: Winston & Strawn Executive Vice President and 1400 L Street, N.W. Chief Operating Officer Washington, DC 20005-3502 Entergy Operations, Inc. P. O. Box 31995 Manager- Licensing Jackson, MS 39286 Entergy Operations, Inc. River Bend Station General Manager- Plant Operations P. O. Box 220 ~ Entergy Operations, Inc. St. Francisville, LA 70775 River Bend Station P. O. Box 220 Senior Resident inspector St. Francisville, LA 70775 P. O. Box 1050 St. Francisville, LA 70775 Director- Nuclear Safety Entergy Operations, Inc. President of West Feliciana River Bend Station Police Jury P. O. Box 220 P. O. Box 1921 St. Francisville, LA 70775 St. Francisville, LA 70775 Vice President- Operations Support h Regional Administrator, Region IV U.S. Nuclear Regulatory Commission Entergy Operations,Inc. P. O. Box 31995 611 Ryan Plaza Drive, Suite 1000 Jackson, MS 39286-1995 Arlington,TX 76011 Attomey General Ms. H. Anne Plettinger State of Louisiana 3456 Villa Rose Drive P. O. Box 94095 i Baton Rouge, LA 70906 Baton Rouge, LA 70804-9095 ; I Administrator Wise, Carter, Child & Caraway i Louisiana Radiation Protection Division P. O. Box 651 i P. O. Box S2135 Jackson, MS 39205 Baton Rouge, LA 70884-2135 ! Mr. Randall K. Edington Vice President- Operations i Entergy Operations, Inc. ! River Bend Station P.O. Box 220 St. Francisville, LA 70775 j O
7590-01-P U. S. NUCt FAR REGULATORY COMMISSION DOCKET NO. 50-458 LICENSE NO. NPF-47 ENTERGY OPERATIONS. INC. I RECEIPT OF PETITION FOR DIRECTOR'S DECISION UNDER 10 CFR 2 206 l Notice is hereby g'iven that by Petition dated September 25,1998, David A. Lochbaum (Petitioner), acting on behalf of the Union of Concemed Scientists (UCS), has requested that )
'the U.S. Nuclear Regulatory Commission (NRC) take action with regard to the River Bend Station (RBS), operated by Entergy Operations, Incorperated. Petitioner requests that enforcement action be taken to require an immediate shutdown of the RBS, and that the facility remain shut down until all failed fuel assemblies are removed from the reactor core. As an attemate action, UCS also stated that following the requested shutdown, RBS could be resterted after its design and licensing bases were updated to permit operation with failed fuel
, sssemblies. Additionally, the Petition requested a public hearing to present new plant-specific l information regarding the operatien of RBS, as well as to discuss a UCS report dated April 2,1908, entitled " Potentia 1 Nuclear Safety Hazard / Reactor Operation With Failed Fuel
,. Cladding."
As the basis for the request, e:amples were cited in the Petition (summarized below) whers, in the Petitioner's opinion, the RBS Updated Safety Analysis Report (USAR) does not j allow for operation with pre-existing fuel failures: (1) Integrity of the fuel barrier is an explicit criterion in addition to radiation requirements, and RBS is violating "the spirit, if not the letter, of [USAR Section 15A, Table 15A.2-4) Criterion 4-2 since thc fuel barrier has already failed, albeit to a limited extent." (2) The USAR description for six design-bases events includes either the statement that the fuel barrier maintains its " integrity and functions as designed," or that "no radioactive material is released from the fuel," as a consequence of the event.
e . I l C- It is the Petitioner's view that the analyses associated with these events
- appear [s] valid only when the River Bend Station is operated with no failed fuel i
- assemblies."
l The Petitioner fWr reasserted the UCS position that nuclear power plants operating l with fuel cladding failures were potentially unsafe and were in violation of Federal regulations. { 1 in its April 1998 report, the UCS stated that it has not been demonstrated that the effects from j design-bases transients and accidents (i.e., hydrodynamic loads, fuel enthalpy changes, etc.) l prevent pre-existing fuel failures from propagating. Therefore, the Petitioner concluded that it was possible that "significantly more radioactive material will be released to the reactor coolant ! system during a transient or accident than that experienced during steady state operation." in addition, the Petitioner also stated that, by operating with possible failed fuel cladding, RBS is ,
\
violating its licensing basis for the radiation worker protection (as low as reasonably achievable [ALARA]) program as it is described in USAR Sections 12.1.1, " Policy Considerations," and 12.1.2.1, " General Design Considerations for ALARA Exposuias." The request is being treated pursuant to 10 CFR 2.206 of the Commission's regulations. The request has been referred to the Director of the Office of Nuclear Reactor Regulation. As provided by Section 2.206, appropriate action will be taken on this petition within a reasonable time. By letter dated October 29,1998, the Director denied Petitioner's request for enforcement action to require Entergy Operations, Inc., to immediately shut down RBS. In addition, the
; Director also extended an offer to the Petitioner for an informal public hearing at a date to be determined. A copy of the petition is available for inspection at the Commission's Public Document Room at 2120 L Street, N.W., Washington, D.C. 20555-0001.
FOR THE NUCLEAR REGULATORY COMMISSION
.Xbtb? ,
Frank J. gliW,McSl3 Director Office of Nuclear Reactor Regulation Dated at Rockville, Maryland, This 29th day of October 1998 l
?: . . .P I)' UNION OF CONCERNED SCIENTISTS September 25,1998 Mr. L. Joseph Callan Executive Director for Operations United States Nue! ear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
PETITION PURSUANT TO 10 CFR 2.206, RIVER BEND STATION
Dear Mr. Callan:
The Union of Concerned Scientists submits this petition pursuant to 10 CFR 2.206 requesting that the River Bend Station be immediately shut down and its operating license suspended or modified until such time that the facility's design and licensing bases are properly updated to permit operation with failed fuel assemblies or until all failed fuel assemblies are removed from the reactor core. fs (j Background On April 2,1998, UCS provided the Nuclear Regulatory Commission with a copy of our report titled
" Potential Nuclear Safety Hazard / Reactor Operation with Failed Fuel Cladding." We concluded:
UCS considers nuclear plants operating with fuel cladding failures to be potentially unsafe and to be violating federal regulations. NRC Daily Event Report No. 34815 dated September 21,1998, provided the following information about an event notification received from the River Bend Station licensee: The licensee notified the Louisiana Department of Environmental Quality of a possiMe defect in fuel cladding. The notification is required by plant procedures, ne possible clad defect was identified by the offgas pretreatment radiation monitor. The monitor is located upstream of offgas treatment equipment and indicated a small increase from 80 to 100 millirem per hour followed by a subsequent rise to about 300 millirem per hour. The level since then has been slowly decrersing. There has been no measur-ble increase in radioactive releases from the plant and radioactive releases remain well below the linus of the technical requirements manual and 10CFR20. Plant personnel are implementing site procedures to address the issue and taking appropriate actions. On September 22,1998, UCS reviewed the latest Updated Final Safety Analysis Report (UFSAR) available in the NRC's Public Document Room and confumed that the generic concerns documented in our April 1998 report appear to apply to the River Bend Station. (l G Washington Office: 1616 P Street NW Suite 310 . Washington DC 200361495 e 202 332-0900 . FAX: 202 332 0905 ( Cambridge Headquarters: Two Brattle Square . Camt: idge MA 02236-9105 . 617-547 5552 . FAX: 617 864 9405 Califomia Office: 2397 Shattuck Avenue Suite 203 . Berkeley CA 947041567 510-843-1872 . FAX: 510-843-3785 EDO -- G980%? ,
V' . . y Page 2 of UFSAR 'Section 15A.2.8, " General Nuclear Safety Operational Criteria," stated: ne plant shall be operated so as to avoid unacceptable consequences. UFSAR Table 15A.2-4," Unacceptable Consequences Criteria Plant Event Category: Design Basis Accidents," defined ' unacceptable consequences
- as follows:
.41 . Radioactive material release exceeding the guideline values of 10CFR100.
42 ' Failure of the fuel barrier as a result of exceeding mechanical or thetmal limits. 4-3 Nuclear sysum r:resses exceeding that allowed for accidents by applicable industry codes. 4-4 Containment stresses exceeding that allowed for accidents by applicable industry codes when containment is required. . 4-5 Overexposure to radiation of plant main control room personnel. The current operating condilon at the River Bend Station apparently violates the spirit, ifnot the letter, of Criterion 4-2 since the fuel barrier h.is already failed, albeit to a limited extent. His UFSAR text does _ ng accept a low level of fuel barrier failure based on meeting the offsite and onsite radiation protection limits. Integrity of the fuel barrier is an explicit criterion in addition to the radiation requirements. UCS reviewed the UFSAR Chapter 15 description of accident analyses performed for the River Bend Station. UFSAR Section 15.1.1.4, " Barrier Performance," for the loss of feedwater heating event stated: De consequences of this event do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure vessel, or containment are designed; therefore, these barriers maintain their integrity and function as designed. UFSAR Sections 15.1.2.4 for the feedwater controller failure - maximum event,15.1.3.4 for the pressure regulator failure - open event, and 15.2.1.4 for the pressure regulator failure - closed event all contain comparable statements that barrier performance was not performed because the fuel remained intact. Dese analyzed even's appear to be valid only when the River Bend Station is operated with no failed fuel assemblies. Operation with pre-existing fuel failures (i.e., the current plant configuration) appear to be outside of the design and licensing bases for these design bases events. UFSAR Section 15.4.2.5, " Radiological Consequences," for the control rod withdrawal error at power event stated: An evaluation af the radiological consequences was not made for this event since no radioactive material is released from the fuel. UFSAR Section l$.4.5.5," Radiological Consequences," for the recirculation flow control failure with increasing flow event stated: ' An evaluation of the radiological consequences is not required for this event since no radioactive material is released from the fuel These analyzed events also appear valid only when the River Bend Station is operated with no failed fuel assemblies. Operation with pre-existing fuel failures (i.e., the current plant configuration) appear to be outside of the design and licensing bases for these design bases events.
y e , w-September 25,1998 Page 3 of 4
- The effect from pre-existing fuel failures was considered, at least partially, for one design bases event.
- UFSAR Section 15.2A.5.1," Fission Product Release from Fuel," for the main steam isolation valve closure event stated:
- While no fuel rods are damaged as a consequence of this event, fission product activity associated with
^
normal coolantactivity levels as well as that released from previously defective rods is released to the suppression pool as a consequence of SRV [ safety relief valve) actuation and vessel depressurization. The aforementioned design bases events (e.g., control rod withdrawal error at power, loss of feedwater heating, et al) are n_oj bound by these results because the radioactive material,is not " scrubbed" by the suppression pool water asitis in the MSIV closure event. As detailed in UCS's April 1998 report on reactor operation with failed fuel cladding, it has not been demonstrated that the effects from design bases transients and accidents (i.e., hydrodynamic loads, fuel enthalpy changes, etc.)
. prevent pre-existing fuel failures from propagating. It is therefore possible that significantly more radioactive material will be released to the reactor coolant system during a transient or accident than that experienced during steady state operation.Thus, the existing design bases accident analyses for River Bend Station do not bound its current operation with known fuel cladding failures.
In addition to operating with non-bounding design bases accident analyses, it appears that the River Bend licensee is also violating its licensing basis for worker radiation protection. UFSAR Section 12.1.1, " Policy Consideration," stated: [ - ne purpose of the ALARA [as low as is reasonably achievable] program is to maintain the radiation exposure ofplant personnel as far below the regulatory limits as is reasonably achievable. UFSAR Section 12.1.2.1," General Design Considerations for ALARA Exposures," stated that River Bend's efforts to maintain in-plar.t radiation exposure as iow as is reasonably achievable included: Minimizing radiation levels in routinely occupied plant areas and in vicinity of plant equipment expected to require the anention of plant personnel. According to NRC Information Notice No. 87 39, " Control of Hot Particle Contamination at Nuclear Plants:" A plant operating with 0.125 percent pin-hole fuel cladding defects showed a five. fold increase in whole-body radiation exposure :ates in some areas of the plant when compared to a sister plant with high integrity fuel (<0.01 percent lenkers). Around certain plant systems the degraded fuel may elevate radiation exposure rates even more. Industry experience ^=== meted that reactor operation with failed fuel cladding increased radiation exposures for plant workers. The River Bend licensee has a licensing basis requirement to maintain radiation exposures for plant workers as low as is reasonably achievable. He River Bend licensee informed the NRC about potential fuel cladding failures. It could shut down the facility and remove the failed fuel assemblies from the reactor core. Instead, it continues to operate the facility with higher radiation levels. Since it appears that operation with one or more failed fuel assemblies is not permitted by its design and licensing bases, River Bend must be immediately shut down. De facility must remain shut down until:
- 0 - De River Bend heensee removes the failed fuel assemblies from the reactor core.
-OR-Q nc River Bend licensee properly updates the plant's design and licensing bases to permit the plant to operate with known fuel damage.
5
- September 25,1998 r Page 4 of 4 i'
Basis for Requested Action .
' UCS is a non profit, public interest organization with sponsors across the United States, including Louisiana. UCS monitors performance at nuclear power plants in the United States against safety regulations promulgated by the NRC to protect the public and plant workers. When real or potential j erosion of mandated safety margins is detected, as is cuttently indicated at this time at River Bend, UCS engages the NRC, the media, and other authorities to resolve the safety concems. {
Requested Actions j UCS petitions the NRC to require the River Bend Station to be immediately shut down and that the facility remain shut down until all of the failed fuel assemblies are removed from the reactor core. ' Altematively, the plant could be restarted after its design and licensing bases were properly updated to reflect continued operation with failed fuel assemblies. { UCS respectfally requests a hearing on this petition to present new information on reactor operation with failed fuel assemblies. This new information will include, but is not limited to, a discussion of the April 1998 UCS report and the plant specific information regarding River Bend. While our concerns apply to River Bend, we respectfully request that this hearing be held in the DC area since the issue affects all operating nuclearpowerplants.
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(') Sincerely, i N JVMNIO David A. Loc baum Nuclear Safety Engineer , 1 enclosure: " Potential Nuclear Safety Hazard / Reactor Operation with Failed Fuel Cladding," April 22,1998 O o
1 UNION OF qU CONCERNED SCIENTISTS Potential Nuclear Safety Hazard Reactor Operation with Failed Fuel Cladding The Union of Concerned Scientists has identified a potential safety hazard at nuclear power l plants that operate with small cracks and holes in the metal tubing, also called cladding, ' containing their fuel. The fuel cladding is a vital barrier between highly radioactive materials and the environment. From a review of available documentation, UCS concludes that federal regulations require this barrier to be intact during plant operation. There is a good reason for these regulations - the public cannot be harmed as long as the fuel cladding remains intact. Ifit is not intact, radioactivity will be released to the plant and the environment. Such a release could I affect the health ofplant workers and members of the public. In addition, fuel rods with degraded cladding may break apart during an accident and prevent safety equipment from functioning. Despite these potentially serious consequences, nuclear plants routinely operate with defective l fuel cladding. In fact, many, if not all, nuclear plaats have operated with damaged fuel cladding. ! UCS recommends that the Nuclear Regulatory Commission (NRC) enforce federal regulations which prohibit nuclear plants from operating with defective fuel cladding. These regulations allow the NRC to permit nuclear plants to operate with defective fuel cla' dding, but only when their owners establish acceptable boundaries based on studies of both normal operating and Q accident conditions. Until these safety concerns are resolved, UCS considers nuclear plants operating with fuel cladding failures to be potentially unsafe and to be violating federal l regulations.
Background
The following sections discuss: design and licensing bases requirements for nuclear plants; their specific application to nuclear finel design; the use of multiple barriers in protecting the public; the role of the fuel cladding as a barrier; the experience with fuel cladding failures, and the potential safety hazards from fuel cladding failures. Desien and Licensine Bases Reauirements Design and licensing bases requirements establish safe operating boundaries which are supported by extensive safety analyses. Operating within the boundaries provides reasonable assurance that the public will be protected if there is an accident. The safety or danger of operating outside the boundaries has not been analyzed. As a result, safety margins may be compromised when boundaries are crossed, increasing the risk to the public. Therefore, federal regulations do not permit plants to operate in unanalyzed conditions. Fuel Desien Nuclear plant are powered by fuel rods which contain uranium dioxide pellets roughly the size and shape of a large pencil eraser stacked within 12 to 14 feet long metal tubes sealed at each p Washington Office: 1616 P Street NW Suite 310 e Washington DC 200361495 e 202 332 0300 e FAX: 202 332-0905 Cambridge Headquarters: Two Brattle Square e Cambridge MA 02238-9105 e 617 547 5552 e FAX: 617-864-9405 Califomia Omce: 2397 Shattuck Avenue Suite 203 Berkeley CA 94704-1567 e 510-8431872
- FAX: 510-843-3785
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, Potential Nuclear Safety Hazard Reactor Operation with Failed Fuel Cladding ^O end with welded metal caps.' A simplified drawing of a fuel rod is shown in Figure 1. The fuel tubes are also called the fuel cladding. Fuel cladding is like the gas tank in a car - if the tank is breached, highly volatile gasoline can spill out to threaten the safety of its passengers and innocent bystanders, as well as degrading the environment. When fuel cladding is breached, highly radioactive material spills out to threaten the safety ofplant workers and the public.
All operating US nuclear power plants use fuel assemblies containing square arrays of fuel rods. A typical fuel assembly is illustrated in Figure 2. As shown in this figure, the fuel rods must remain intact to provide the overall structural integrity of the fuel assemblies. The fuel design bases ensure that "the fuel is not damaged as a result of normal operation and anticipated operational occurrences." 2 The phrase "not damaged," as used by both the NRC and nuclear j plant owners, means that the fuel rods are not damaged to the point where they would fail.3 Thus, the fuel design bases includes the explicit requirement that fuel cladding remains intact during normal operation. Defense-in-Depth Barriers The splitting, or fissioning, of uranium atoms in the fuel rods releases energy that heats water - i nuclear energy that powers the plant. Byproducts of the fission process include radioactive gases and solids. Plutonium is also produced h the nuclear reactions. These radioactive materials emit gamma rays along with alpha and beta particles which can cause damage to the human body. The fuel cladding keeps the radioactive materials contained. If the cladding is defective, radioactive i h materials will leak into the water which surrounds the cladding and keeps the fuel rods cooled. Q This water is contained within the reactor vessel and the piping connected to it, which form a second barrier to contain the radioactive materials. If the piping fails, contaminated water spills ; into the reactor containment building. The reactor vessel and its piping are located within a l
. reactor containment building w'hich fonns a third barrier. Because the reactor containment building is not leak tight, it reduces, but does not eliminate, the possibility that radioactive material would escape. Figure 3 shows a simplified drawing of these three barriers.
Three barriers between the radioactive material and the environment imply that one barrier can be breached during plant operation leaving two intact barriers to protect the public. However, the safety analyses assume that all three barriers are intact prior to any accident. Let's assume the
- rupture of a pipe connected to the reactor vessel breaches one of the barriers. If the pipe rupture occurs when the fuel cladding is defective, then two of the barriers are breached. The remaining barrier, the reactor containment building, only reduces the amount of radioactive material released to the environment. Thus, all three barriers must be intact during plant operation for the public to be protected.
3 Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.3.2.1," Fuel Rod Mechanical Design," and General Electric Company, " Licensing Topical Report / General Electric Standard Application for Reactor Fuel," NEDO-240ll-A-4, January 1982. 8 Nuclear Regulatory Commission, NUREG-0800. Standard Review Plan, Section 4.2, Fuel System Design.
' Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2, Fuel System Design,
.O_ GPU Nuclear Corporation, Oyster Creek Nuclear Generating Station Updated Final Safety Analysis Report, Section 4.4.2," Description of Thermal and Hydraulic Design of the Reactor Core."
April 2,1998 Page 2 4
,- Potential Nuclear Safety Hazard Reactor Operation with Failed Fuel Cladding b,_ The fuel cladding is the most important of the three barriers. If the fuel cladding remains intact, the other two barriers . completely fail and the public will still be protected. The intact fuel ; cladding contains the radoactive gases and solids and prevents them from being released to the atmosphere. The public cannot be harmed from a nuclear plant accident in which the fuel l cladding remains intact. But, as the next section indicates, nuclear plants routinely operate with this vital barrier seriously degraded. f Fuel Claddine Failure Exnerience Numerous fuel cladding failures from various causes have been reported over the years. For example, the water flowing thrmgh the reactor core has caused fuel rods to sway back and forth. In this situation, the fuel rods vibrate against the grid (shown in Figure 2) and damage the cladding. At other plants, debris in the reactor water, such as metal flakes from rusted piping, has lodged against the grid. The friction from the vibration of this debris damaged the cladding. Another failure mode results when fuel pellets expand faster than the fuel rod cladding (see I Figure 1) as their temperatures increase. The expanding pellets stretch the cladding, sometimes until it cracks or splits. Finally, the welds holding the upper and lower end plugs to the fuel rod cladding (see Figure 1) have sometimes been defective, causing pinhole leaks or even cracks to form. Other failure modes have been experienced too. Many, if not all, nuclear plants have experienced fuel e adding failures during their lifetimes. Few plants have shut down early to remove failed fuel rods. Leaking fuel rods are detected by increased radioactivity levels in the reactor vessel's liquid and gaseous releases.' Not surprisingly, the radioactivity levels rise significantly when fuel cladding , fails. The causes of fuel cladding failures cannot be determined until the plant is shut down and l the leaking fuel rods examined. The following reports illustrate recent fuel cladding failure incidents and include some serious events. The Vermont Yankee plant recently operated with at least one failed fuel rod for many months.5 Its owners elected to operate with the leaker (s) until the plant's next scheduled refueling outage l in the spring of 1998 rather than incur the cost of an unscheduled shut down.' The Brunswick Unit 1 plant.in North Carolina operated during 1997 with fuel cladding failures that its owners , tolerated.' The Surry plant in Virginia also operated in 1997 with failed fuel cladding.' These ! incidents demonstrate that nuclear plants continue to operate with fuel cladding failures. l
' Entergy Operations, River Bend Station 1. pdated Final Safety Analysis Report, Section 4.2.4.2,"Online Fuel System Monitoring,"and Section i1.5.2.2. I," Main Steam Line Radiation Monitoring System."
8 Nuclear Regulatory Commission, Daily Event Report, DER No. 33152, October 28,1997.
- Vermont Yankee Nuclear Power Corporation, Presentation to Vermont State Nuclear Advisory Panel, December 3, 1997.
' Johan Blok and Roger Asay, Centec XXI," Pinpoint fuel leaks to improve nuclear economics," Power, January / February 1998.
April 2,1998 Page 3 l l 1
Potential Nuclear Safety Hazard D Reactor Operation with Failed Fuel Cladding U ! A few years ago, the owner of the Point Beach Nuclear Plant in Wisconsin reponed a significant event in which "The fuel cladding was failed to the extent that fuel pellets could be seen through j the hole in the clad. However, no pellets escaped from the rod." The fuel rod failure was detected I when the radioactivity levels of the reactor water rose to a level that was "10 percent of that allowed by (Point Beach Nuclear Plant's operating license]."' In other words, the plant's ) operating license would have allowed it to remain running with up to nine other similarly failed i fuel rods. This event suggests that the restrictions on reactor water radioactivity levels are too j high to pr event operation with gaping holes in fuel rod cladding. ' At the Palisades plant in Michigan, three portions of a broken fuel rod were discovered in different parts of the reactor. One segment, nearly 5% feet long, was missing about one third of its fuel pellets. A second segment,4% feet long, and a third segment,1% feet long, appeared to contain all their fuel pellets. This event is disturbing because it highlights how fragile the cladding can become during normal operation. At Palisades, this fuel rod literally fell apan as it was being removed from the reactor core and radioactive material was lost, including highly toxic plutoniunt. Fuel Cladding Failure Conseauences What is the safety threat from a nuclear plant operating with fuel cladding failures? The fact that many plants have operated for many years with failed fuel cladding could be taken to imply an 0 acceptable safety record. However, that is not the case. That fact demonstrates, at most, that the public is protected with fuel cladding failures during normal plant operation. It does not provide any reason to believe that the public will be protected in the event of an accident. It also does not provide any reason to believe that nuclear workers will be prMected during normal plant operation with failed fuel cladding. What might happen if a nuclear plant with failed fuel cladding had an accident? A common accident scenario involves breaking a large pipe connected to the reactor vessel. Water and steam rush out of the reactor vessel through the broken pipe. The water flow in the reactor core, instead of flowing from the bottoms of the fuel assemblies to their tops, may flow across the fuel assemblies. This cross flow ' pushes' the fuel rods to the side rather than towards the top. Cladding that is weakened may fail under this side force. The plant's response to the pipe break is to shut down. Control rods are automatically inserted into the reactor core to stop the fissioning process. Fuel rods which fail and shift out of their vertical alignment may prevent the insertion of control rods. The safety analyses assume that the control rods can be inserted and shut down the reactor. Can fuel cladding failures cause such problems during this accident scenario? No one knows. Pre existing fuel cladding failures have not been considered in the
- Nuclear Regulatory Commissioa, inspection Report 50-280/97-10, December 15,1997.
' Wisconsin Electric Power Company, Licensee Event Report No. 85 002-01," Failed Fuel Rod in Assembly H14 Point Beach Nuclear Plant Unit 1," May 19,1986.
United States Nucirr Regulatory Commission. Information Notice 93-82, "Recent Fuel And Core Performance Problems In Operating Reactors," October 12,1993.
April 2,1998 - Page 4
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, Pctantial Nuclear Safety Hazard Reactor Operation with Failed Fuel Cladding O' safety analyses for this accident or any other accident. Yet, nuclear plants routinely operate with 1 such fuel cladding failures. What happens if fuel cladding failures increase the severity of nuclear plant accidents? Since plant safety analyses assume that fuel cladding is undamaged when accidents occur, the failures may cause more radioactivity to be released to the environment than has been previously considered. After all, a key barrier confining this highly radioactive material is already breached when the accident begins. Under no circumstances will less radioactivity be released. Thus, it is imperative from a public health standpoint that nuclear plants do not operate with fael cladding failures unless safety analyses are performed which demonstrate that the consequences from accidents under these co. ditions are acceptable. 1 Summary The fuel cladding is the most important of the three barriers between highly radioactive material t and the environment. As long as the fuel cladding remains intact, no nuclear plant accident can I threaten public health and safety. Yet, nuclear plants routinely operate with damaged fuel cladding.
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Safety analyses assume that the fuel cladding is intact when accident scenarios begin. Operation with pre-existing fuel cladding failures may mean that a nuclear accident will have more severe consequences than predicted by the invalidated safety analyses. Thus, UCS considers a nuclear p plant operating with defective fuel cladding to represent an increased risk to the public.
'd l l
The fuel design bases require the fuel cladding to remain intact during normal plant operation. l Federal safety regulations require that plants operate within the boundaries established by their i design bases. Therefore, UCS concludes that operating a nuclear plant with failed fuel cladding i violates federal safety regulations. See Attachment I for details of UCS's assessment of reactor operation with failed fuel cladding. ALARAIssue Nuclear plant owners are required by federal regulations to keep the release of radioactive materials "as low as reasonably achievable" (ALARA)." According to the NRC, "a plant operating with 0.125 percent pin hole fuel cladding defects sh wed a general five-fold increase in whole-body radiation exposure rates in some areas of the plant when compared to a sister plant with high-integrity fuel (<0.01 percent leakers). Around certain plant systems the degraded fuel may elevate radiation exposure rates even more."i2 The " sister plants" were virtually identical because they were built at the same time by the same owner on the same site. The
" Title 10 of the Code of Federal Regulations, Sections 50.34a, " Design objectives for equipment to control releases of radioactive material in efiluents - nuclear power t eactors," and 50.36, " Technical specifications," and Title 10 of the Code of Federal Regulations, Part 50, Appendix I," Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Reasonably Achievable" for Radioactive Material in Light Water-Cooled Nuclear Power Reactor Effluents."
" United States Nuclear Regulatory Commission, Information Notice No. 87-39, " Control Of Hot Particle b Contamination At Nuclear plants," August 21,1987.
April 2,1998 Page 5
,' Potential Nuclear Safety Hazard n Reactor Operation with Failed Fuel Cladding U
significant variation in radiation exposure rates is nnl due to thicker concrete or other design differences - it is due to the failed fuel cladding. UCS is troubled by this NRC evidence because it shows a significantly increased risk to nuclear plant workers at a facility operating with just 0.125 percent fuel cladding failures. Many plants consider it permissibl: to operate with eight times as many fuel cladding failures (up to 1.0% failures). Fuel cladding defects release radioactive materials into the reactor water. The water carries them to all parts of the plant, contaminating eqaipment throughout the facility. Workers conducting equipment inspections and maintenance receive higher radiation exposures. Indeed, some plant wkers have received radiation doses far greater than allowed b highly radioactive material released through fuel cladding defects." y federal reg It is a well-documented fact that plant operation with defective fuel cladding cignificantly increases personnel exposures. Federal regulations .equires nuclear plant owners to keep the release of radioactive materials as low as reasonably schievabic. Thenfore, it is both an illegal activity and a serious health hazard for nuclear plants t-) continue operatinE with fuel cladding damage. Conclusions And Recommendations Conclusions
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r It is UCS's considered ophion that existing design and licensing requirements do not allow plants to operate with known fuel cladding failures. In addition, federal regulations require formal NRC approval prior to any nuclear plant operating with fuel cladding failures. Such approval has neither been sought nor granted. UCS's evaluation (see attachment 1) suggests that both the probability and consequences of postulated accidents may be increased when nuclear plants operate with pre-existing fuel cladding failures. Thus, operation with fuel cladding failures is a violation of federal regulations which represents a potential threat to public health and safety. UCS's assessment was generic. Consequently, this conclusion does not explicitly apply to any operating plant. However, UCS's assessment identified the strong potential for operation with fuel cladding failures to be an illegal activity unless the plant's owners performed a plant-specific safety evaluation which established such operation as acceptable and the NRC has formally reviewed and approved this safety evaluation. Absent both of these conditions, it seems highly probable that any plant operating with fuel . 3dding failures is violating its design and licensing 1ases requirements, a condition not allowed by. federal safety regulations. It further appears that such illegal operation may br~e serious safety implications. Finally, operation with fuel cladding damage also seems to violate the AI ARA concept mandated by federal regulations, thus exposing plant workers to undue risk.
" United Smes Nuclear Regulatory Commission, Information Notice b ' 39,"Contml Of Hot Particle Contamination At Nuclearplants," August 21,1987.
April 2,1998 Page 6
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, Potential Nuclear Safety Hazard Reactor Operation with Failed Fuel Clad < ling .O UCS's renarch for this assessment did not locate any information which suggests that operation with failed fuel cladding has been previously evaluated pursuant to federal regulations. There is considerable documentation on fuel cladding failure events, on inspections of failed fuel rods, and on various fuel damage mechanisms. Despite extensive, focused efforts, UCS was unable to find any indication that the safety implications of plant operation with failed fuel cladding have been considered by the fuel vendors, the NRC, or nuclear plant owners. This non-existent data further reinforces UCS's conclusions that operation with failed fuel cladding has not been properly analyzed by the industry, has not been approved by the NRC, and is both potentially unsafe and illegal. Recommendations UCS recommends that the Nuclear Regulatory Commission take appropriate steps to prohibit nuclear power plants from cperating with fuel cladding damage until the safety concems raised in this report are resolved. These appropriate steps include, but are not limited to, the following: e Plant owners should be required to shut down their facilities upon detection of a fuel cladding failure. The plants must not restart until the failed fuel rods are removed. Plant owners should be required to evaluate the safety implications of operating with failed fuel cladding in accordance with federal regulations. If these safety evaluations are unable to justify continued operation, the plants should be shut down. For the long term resolution of the safety concems raised in this report, UCS recommends that the Updated Final Safety Analysis Reports (UFSARs) be revised. These revisions would establish safe boundaries for operation. After these boundaries are drawn and incorporated into the UFSARs, plants could continue to operate with failed fuel cladding as long as the failures remained within the previously analyzed region. If the amount of failed fuel cladding exceeded the boundarie, then the plant should face the options recommended above. I N O April 2,1998 Page 7
Attachment 1 Unreviewed Safety Question Assessment Unreviewed Safety Question Assessment This attachment contains UCS's evaluation for reactor operation with failed fuel cladding. Our evaluation applied federal regulations for determining when a proposed mode of operation crosses the plant's authorized boundaries and thus requires prior NRC approval. As the results clearly indicate, reactor operation with failed fuel cladding requires NRC approval. Yet, such approval has neither been sought nor granted. The NRC issues an operating license for a nuclear power plant after reviewing its design and procedures. The plant's owners may modify the facility and revise its procedures as long as the changes do not alter the bases for the NRC's approval of the operating license. A change which alters the operating license bases is called an unreviewed safety question (USQ). For example, a proposed change that reduces the plant's safety margin is an unreviewed safety question because the NRC may have relied on the greater margin in granting the plant's operating license. Likewise, a proposed change that maintains the existing safety margin but does so by operator actions instead of automatic equipment operation is also an USQ because the NRC's approval may have relied on the automatic protective features. When a proposed change involves an USQ, NRC approval must be obtained in advance. Federal regulations specify that a proposed change involves an USQ if: (1) the probability of occurrence or the consequences of an accident or malfunction of equipment impodant to safety previously evaluated in the safety analysis report A) ( may be increased; or (2) a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (3) the margin of safety as defined in the basis for any technical specification is red 6ced." Federal regulations require nuclear plant owners to obtain NRC permission prior to conducting any activity for which the answer to one or more of these questions is anything but "NO." As UCS's nuclear safety engineer, I reviewed publicly available documentation to determine if these criteria are satisfied for plants operating with fuel cladding failures. Prior to joining UCS, I worked in the nuclear industry for over 17 years where I developed, reviewed, and assessed literally thousands of USQ determinations. I divided the first criterion above into the " probability" and " consequences" elements for clarity. The scope of this evaluation was limited to four types of documentation: 1) the Updated Final ' Safety Analysis Reports (UFSARs) for four of UCS's focus plants (the Calvert Cliffs plant in Maryland, the Oyster Creek plant in New Jersey, the River Bend plant in Louisiana, and the Millstone Unit 3 plant in Connecticut); 2) the non proprietary version of the fuel design topical report submitted by a vendor (General Electric); 3) the standard technical specifications prepared by all four reactor manufacturers (Westinghouse, General Electric, Babcock & Wilcox, and
' " Title 10. " Energy," of the Code of Federal Regulations, Section 50.59," Changes, tests and experiments," l April 2,1998 Page 8
' Attachment 1 l Unreviewed Safety Question Assessment !
Q. Combustion Engineering); and 4) NRC correspondence on fuel cladding failure events. The l l
, results from this evaluation follow.
e Criterion la: Msy the probability of occurrence of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report be increased by operation with failed fuel cladding? The standard technical specifications prepared by Westinghouse, General Electric, Combustion Engineering, and Babcock & Wilcox (vendors for all of the plants operating in the United I States).specify that "The fuel cladding must not sustain damage as a result of normal ' operation."'5 The NRC considers fuel cladding to be damaged when its integrity is lost." The detection of fission products outside the fuel rods is irrefutable evidence that fuel cladding integrity has been lost.
' The standard technical specifications are the templates from which individual plant operating licenses were derived. Since these specifications establish zero defects as the minimally acceptable standard,' operation with fuel cladding failures increases the probability of
" malfunction of equipment important to safety," namely the fuel itself, to 100%. For this reason alone, the answer to this question is XES.
To apply the above generic assessment to a specific plant, UCS looked at available ! (] v documentation for the Oyster Creek Nuclear Generating Station in New Jersey. A design basis , for Oyster Creek is "to ensure that no fuel damage will occur in normal operation or operational transients caused by reasonable expected single operator error or equipment malfunction."" Fuel
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tod damage "is dc5ned as a perforation of the cladding which would permit the release of fission l product to the reactor coolant." Thus, the detection of failed fuel rod (s) at Oyster Creek would ; be an equipment malfunction placing the plant outside its design basis. Again, the answer to thi: l question is XES. ' A fuel cladding defect may allow gases within a fuel rod to leak out. A defect may also allow water to leak in. It appears that leakage in either direction may also increase the probability that the fuel cladding will not perform its necessary safety function.
" Babcock & Wilcox Company, Standard Technical Specifications, Section B 2.1.1.," Reactor Core SLs,"
Combustion Engineering, Standard Technical Specifications, Section B 2.1.1," Reactor Core SLs," General Electric Company, BWR/4 Standard Technical Specifications, Section B 2.1.1, " Reactor Core SLs," and Westinghouse Electric Corporation, Standard Technical Specifications, Section B 2.1.1," Reactor Core SLs."
" Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2, " Fuel System Design."
" GPU Nuclear Corporation, Oyster Creek Nuclear Generating Station Updated Final Safety Analysis Report, Section 4.4.1,"[ Thermal and Hydraulic Design) Design Basis."
, " GPU Nuclear Corporation, Oyster Creek Nuclear Generating Station Updated Final SaTety Analysis Report, Section 4.4.2," Description of Thermal and Hydraulic Design of the Reactor Core."
April 2,1998 Page 9
' Attachment 1 Unreviewed Safety Question Assessment O A fuel cladding defect which allows gases to leak out of a fuel red has at least two potentially adverse consequences. The fuel rods are pressurized with helium during their fabrication to minimize a problem called cladding creep-collapse. The pressure inside a 6uclear plant ranges from 960 to 2,100 pounds per square 9:h at full power. The difference between a fuel rod's external pressure and internal pressure can exert suffcient inward force to cause the cladding to fill tne gaps between fuel pellets." The stress on the cladding can cause it to break. The leakage of helium from a fuel rod reduces its intemal pressure, thus potentially increasing the probability of fuel rod damage from cladding creep-collapse. Inadequate cooling of the fuel is another potential consequence from gases leaking out of a fuel rod. Helium is used to pressurize fuel rods because of its high thermal conductivity.2 The leakage of helium through a fuel cladding defect may slow down the transfer of heat from the fuel to the water. When heat cannot be dissipated from the fuel as quickly as assumed, the fuel temperature will increase and may reach the point at which it begins to melt. The leakage of helium from a fuel rod may reduce heat transfer rates, thus potentially increasing the probability that the fuel is seriously damaged during a loss-of-coolant accident. A fuel cladding defect which allows water to leak into a fuel rod also has at least two potentiaUy adverse consequences. During plant operation, high fuel temperatures prevent water from leaking in through a cladding defect. However, water can enter defects when the plant is shut down and cause fuel rods to become waterlogged. If the plant increases power quickly, the rising (] V fuel temperature may cause the water inside the fuel rods to evaporate and perhaps even boil. The water vapor and steam produced inside the fuel rods, unless it is able to leak out through the defects, increase their pressure. This pressure buildup is suspected to have caused the " bursting" of fuel rods at the Point Beach plant in Wisconsin. Sections of the cladding and sev pellets could not be located when the damaged assemblies were later inspected. ' There is another potential adverse consequence from water leaking into fuel rods. The high operating temperature dissociates the water into hydrogen and oxygen gases. The hydrogen gas interacts with the cladding to form blisters. The blisters embrittle the cladding, leading to perforations.22 To minimize the moisture content, the fuel pellets are dried prior to being !caded
" Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.7.1.1.a," Clad Creepdown/ Creep-Collapse."
2' Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report Section 3.3.2.1, " Fuel Rod Mechanical Design." 2' B. Siegel, Nuclear Regulatory Commission,' " Evaluation of the Behzucc of Waterlogged Fuel Rod Failures in LWRs," NUREG-0303, March 1978. 22 g ) Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.7.2.1," Burnable Poison Rod Design Evaluation." April 2,1998 Page 10
' Attachment 1 Unreviewed Safety Question Assessment into the fuel rods.23 Thus, water leaking into a fuel rod may increase the probability that fuel cladding suffers this type of damage, which is called hydriding.
In fact, failure propagation due to hydriding has already been identified. Recent inspections of failed fuel rods at the Salem plant in New Jersey, the Beaver Valley plant in Pennsylvania, and the Wolf Creek plant in Kansas revealed that, "In some of the affected assemblies, secondarv hydriding also was evident."2' A fuel rod at the Perry Nuclear plant in Ohio experienced a cladding crack measuring 20 inches long, or nearly 13% of the fuel rod's length, caused by secondary hydriding.25 In these events, the initial fuel cladding failures were caused by other mechanisms. These failures later propagated due to hydriding. Thus, operation with fuel cladding failures has the potential for increasing the probability that an imponant barrier protecting the public, namely the fuel cladding itself, fails to adequately confine radioactive materials during a postulated accident. The fuel cladding is considered
" equipment imponant to safety." A fuel cladding failure is therefore a malfunction of equipment important to safety. For this reason, too, the answer to this criterion is XES.
Finally, the NRC's Standard Review Plan states that the fuel design bases ensure that " fuel damage is never so severe as to prevent control rod insenion when it is required." 26 Nuclear plant operation with failed fuel cladding has caused individual fuel rods to break into segments during fuel handling evolutions. If degraded fuel cladding were to similarly break during an accident, the fuel rod segments might interfere with control rod insertion. Thus, for this additional reason, the answer to this criterion is XES.
'e Criterion Ib: May the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis repon be increased by operation with failed fuel cladding?
The NRC reported that the nuclear fuel's design bases are intended to " provide assurance that the fuel system is not damaged as a result of normal operation. 'Not damaged,' as used in the above statement, means that fuel rods do not fail. Fuel rod failure is defined as the loss of fuel rod [ integrity]."27 Thus, the fuel system, including the fuel cladding, must remain undamaged during normal operation. 23 Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.3.2.1, " Fuel Rod Mechanical Design, and Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2," Fuel System Design."
United States Nuclear Regulatory Commission, Information Notice 93-82,"Recent Fuel And Core Performance Problems in Operating Reactors," October 12,1993.
28 United States Nuclear Regulatory Commission, Information Notice 93-82,"Recent Fuel And Core Performance Problems in Operating Reactors," October 12,1993.
Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2, Fuel System Design. .. 27 Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2," Fuel System Design."
April 2,1998 Page11
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' ~ . Attachment 1 Unreviewed Safety Question Assessment O
The safety analysis for the recirculatior. ' low control failure with increasing flow event 2s at the River Bend Station in Louisiana cone" .ded that "An evaluation of the radiological consequences is not required for this event since no radioactive material is released from the fuel."2 If this event were to occur with pre-existing fuel cladding failures, this analysis would be rendered invalid. Since this analysis assumes that the fuel cladding remains intact, its conclusions are invalidated when there are fuel cladding failure ' The safety analysis for the feedwater controller failure maximum demand event20 at River Bend concludes that fuel and pressure vessel " barriers maintain their integrity and function as designed." Obviously, this analysis's conclusion is invalidated when the plant operates with pre-existing fuel cladding failures. 32 The safety analysis for the rod withdrawal error event at River Bend specifies that "An evaluation of the barrier performance was not made for this event since this is a localized event with very little change in the gross core characteristics."" Fuel cladding damage is a localized event. The failed fuel rod has a pinhole leak or a hairline split in its cladding or a cracked v cid at its end cap. If the rod withdrawal error occurs in the vicinity of the fuel cladding defec'., the big change in local characteristics could propagate that defect. Thus, this analysis's conclusion is invalidated when the plant operates with a' fuel rod defect. (^s The safety analysis for a control element assembly ejection event3 ' at the Calvert Cliffs Nuclear v Plant concluded that "the site boundary [ radiological] dose guidelines will be approached." 35
" This potential accident is comparable to a mistake using a bellows to flame a wood fire. If too much air is supplied, the fire may blaze up out of control. Likewise, putting too much water through the River Bend reactor core can cause it to run out of control.
" Entergy Operations, River Bend Station Updated Final Safety Analysis Report, Section 15.4.5.5, "[ Recircul Flow Control Failure with increasing Flow) Radiological Consequences."
This potential accident is similar to the recirculation flow control failure with increasing flow event in that too much water to the reactor core results in an uncontrolled power increase.
Entergy Operations, River Bend Station Updated Final Safety Analysis Report, Section 15.1.2.4,"[Feedwater Controller Failure Maximum Demand) Barrier Performance."
32 This potential accident involves the inadvertent withdrawal of a control rod causing the power produced by the adjacent fuel assemblies ;o increase significantly.
" Entergy Operations, River Bend Station Updated Final Safety Analysis Report, ..., Section 15 4 2 4 "[ Rod Withdrawal Error) Barrier Performance."
" This potential accident is comparable to car engine throwing one ofits pistons. The piston may break the eng i casing. Likewise, the ejected control element assembly rnay break the reactor coolant pressure boundary and allow reactor water to leak out. l 0 38 Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 14.13.2," Sequence of Events [ Control Element Assembly Ejection)."
5 i April 2,1998 Page 12
7 ,' Attachment 1 j Unreviewed Safety Question Assessment l O The analysis found the postulated event acceptable because the plant's design features "will l prevent fuel clad failure, will prevent exceeding the [ reactor coolant system] Pressure Upset l Limit, and will therefore limit the radiological site boundary dose [i.e., the radiation levels ' experienced by a member of the public at the plant's fence) to below the criteria in 10 CFR 100 guidelines." 3 Since this analysis assumes that fuel claddin5 failures are preverned, its conclusions are invalidated when there are pre-existing fuel cladding failures. l The NRC's Standard Review Plan states that the fuel design bases ensure that "the number of fuel rod failures is not underestimated for postulated accidents." 37 Yet, the previous accident analyses underestimated the number of fuel rod failures if those plants operated with fuel cladding failures. Thus, the answer to this criterion is XES. The Wolf Creek plant recently experienced fuel cladding failures affecting 44 fuel rods in three fuel assemblies. According to an NRC report on the problem, "The most severely degraded fuel rod fragmented into three segments during fuel handling operations while offloading the core."'8 Fuel handling operations include removing a fuel assembly from the reactor core, placing it in a device called an upender, lowering the assembly to a horizontal position, transferring it through the reactor containment wall into the fuel handling building, raising the assembly to a vertical position, and moving it to a storage location in the spent fuel pool. These manipulations put dead load force (i.e., gravity) on the fuel assembly and it' sfuel rods. Fuel assemblies are designed to withstand the force associated with these handling evolutions, at least when their fuel cladding is O undamaged. Apparently at Wolf Creek, the force ofgravity was sufficient to cause the structural failure of a fuel rod with previously damaged cladding. ; What if an accident occurred whet. the fuel assemblies with the damaged cladding still resided in l the reactor core? For example, consider the hydrodynamic forces inside the reactor vessel i following a break of a large pipe connected to it. The high energy water escaping through the j break exerts considerable force. The side force on the fuel rods may approach, or even exceed, I the dead load iorce during fuel handling. The weakened fuel cladding may experience structural failure as was encountered during fuel handling. Fuel rod structural failure could have very serious consequences during an accident. The dislodged fuel rod segments could interfere with the insertion of control elements attempting to shut down the reactor. Fuel assemblies are tightly i packed into the reactor vessel. The clearance between fuel assemblies and control elements is fractions of an inch at most. Fuel rod segments would not have to move much in order to interfere with control elements. Thus,' the consequences of previously analyzed accidents could be increased by operation with fuel cladding failures. The answer to this criterion is XES. 8' Baltimore Gas & Electric Company, Calven Cliffs Nuclear Plant Updated Final Safety Analysis Report Se 14.13.4," Conclusion [ Control Element Assembly Ejection]."
" Nuclear Regulatory Commission, NUREG-0800, Standard Review Man, Section 4.2, Fuel System Design.
8' United States Nuclear Regulatory Commission, Information Notice 93 82,"Recent Fuel And Core Perfonnance Problems in Operating Reactors," October 12,1993. April 2,- 1998 Page 13
,' Attachment i unreviewed Safety Question Assessment O e Criterion 2: May the possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report be created by operation with failed fuel cladding? -
. l After residing in the reactor core for one or more cycles of operation, fuel assemblies are moved to the spent fuel pools. " Spent" fuel assemblies continue to generate considerable amounts of.
heat and release deadly amounts of radiation for many years. The worst-case spent fuel pool l accident is typically assumed to be a fuel handling event. The analysis for this event assumes that ; a fuel assembly is dropped onto another fuel assembly." Fuel rods in both assemblies are ' assumed to fail to evaluate the radiological consequences of the event. The spent fuel pools are also analyzed for possible damage resulting from an earthquake. These analyses generally assume that no fuel damage occurs as long as the fuel storage racks remain structurally intact. Some spent fuel pool accident analyses take credit for operation of the spent fuel building's ventilation system. This system routes the building's exhaust air through filters, thus lowering the radiological dose to the public. At many plants, the ventilation system only performs this safety function when fuel handling operations are underway. Spent fuel assemblies with cladding failures may have those failures propagate when subjected to earthquake forces. Radioactive gases released from spent fuel assemblies following an earthquake may cause radiological consequences which exceed those for the fuel handing event if(a) the inventory from more than the fuel rods in two assemblies is released, or (b) credit is taken in the fuel handling e. ' analysis for operation of the spent fuel building's ventilation system but the system is unava ble. Consequently, the answer to this criterion is MAYBE. Criterion 3: May the margin of safety as defined in the basis for any technical specification be reduced by operation with failed fuel cladding? The standard technical specifications prepared by Westinghouse, General Electric, Combustion Engineering, and Babcock & Wilcox (venders for all of the operating plants in the United States) specify that "The fuel cladding must not sustain damage as a result of normal operation and [ anticipated operational occurrences]."" The NRC considers fuel clad mg to be damaged when hs integrity is lost." The detection of fission products outside the fuel rods is irrefutable evidence that fuel cladding integrity has been lost.
" Baltimore Gas & Electric Company, Calven Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 14.18.2," Method of Analysis [ Fuel Handling Accident]."
** Babcock & Wilcox Company, Standard Technical Specifications, Section B 2.1.1., " Reactor Core SLs,"
Combustion Engineering, Standard Technical Specifications, Section B 2.1.1," Reactor Core SLs," General Electric Company, BWR/4 Standard Technical Specifications, Section B 2.1.1, " Reactor Core SLs," and Westinghouse g Electric Corporation, Standard Technical Specifications, Section B 2.1.1, " Reactor Core SLs." V
Nuclear Regulatory Commission NUREG-0800, Standard Review Plan, Section 4.2," Fuel System Design."
April 2,1998 Page 14
' Attrchment 1 Unreviewed Safety Question Assessment The standard technical specifications are the templates from which individual plant ope licenses are derived. Since these specifications establish zero defects as the minimally a standard, operation with fuel cladding failures clearly represents a safety margin reduction. Consequently, the answer to this question appears is.YES. l Conclusion Federal regulations specify that an unrev.iewed safety question is indicated when the answer to any one of the criteria is non-negative. UCS's assessment determined that none of the answers is negative. Three of the answers are unequivocally YES and a fourth is MAYBE. Thus, nuclear power plant operation with failed fuel cladding is clearly an unreviewed safety question. NRC approval is required for a plant to continue operating with fuel cladding failures, n Performed by: ) f//ztO _ A opet.98 David M Lochbaum Nuclear Safety Engineer O April 2,1998 Page 15
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( . 6 UNION OF Reactor Operation with Failed Fuel: U CONCERNED Bliss Reduction Program Needed SCIENTISTS Good afternoon. My name is David Lochbaum. I have been the Nuclear Safety Engineer for the Union of Concerned Scientists (UCS) since October of 1996. Prior to joining UCS I worked in the nuclear industry for more than 17 years, primarily as a consultant. As a consultant, I worked on assignments at the Perry plant in 1995 and 1996. For one of these assignments, I developed a lesson plan on design and licensing bases requirements and presented it to managers, supervisors, and staffin the Design Engineering department. I was a reactor engineer for a total of nearly 8 years at the Hatch, Browns Ferry, Grand Gulf, and Hope Creek nuclear plants - all boiling water reactors similar in design to the River Bend and Perry plants. Among other reactor engineering duties, I was responsible for the fuel integrity monitoring program and updating the accident analyses for each operating cycle. I 1 I am here today because UCS submitted two petitions to the NRC last year. Last spring, we provided the i NRC with our technical report documenting our concerns with nuclear plants that operate with damaged ) fuel. At that time, we were not aware of any plants operating with damaged fuel. But plants had operated j with damaged fuel in the past and were likely tc operate with damaged fuel in the future. Five months later, we learned that the River Bend plant was operating with damaged fuel leaking radiation. Other than a letter from the NRC acknowledging receipt of our technical report, we had not heard from the agency. We still don't know if they agreed with, disagreed with, or even understood the concerns. l I In an effort to prompt the NRC to take some action on our safety concerns, we submitted a petition (q) asking that the River Bend plant be immediately shut down until the leaking fuel was removed or until the plant's owners had performed an analysis showing that it was safe to operate with the leaking fuel. A few weeks later, we teamed that the Perry plant was also operating with failed fuel. So, l we submitted a similar petition. It is our intention to continue illing petitions when plants operate with leaking fuel until the NRC addresses our concerns or until we run out of postr.ge stamps. And we have plenty of stamps. The River Bend plant, located north of Baton Rouge, Louisiana, was granted an operating license by the NRC on November 20,1985. The Perry plant, located northeast of Cleveland, Ohio, was issued an operating license about a year later on November 13,1986. In each case, the NRC issued the operating license after a lengthy, deliberative process through which it concluded there was reasonable assurance that two criteria were met:
- 1. That the facility's design met all applicable regulatory requirements.
- 2. That the facility would be operated and maintained in accordance with all applicable regulatory requirements.
By letters dated September 25,1998, and November 9,1998, UCS petitioned the NRC to require the immediate shut down of the River Bend and Perry plants because they continue to operate even though some of their nuclear fuel is leaking radioactivity - a condition that violates the terms of their license. As a result of this illegal activity, plant workers face greater risk from radiation overexposure during normal plant operation. In addition, both plant workers and the general public may face greater risk from radiation overexposure if an accident were to occur. Washington Office: 1616 P Street NW Suite 310 e Washington DC 20036-1495 202-332-0900 . FAX: 202-332-0905 Cambridge Headquarters: Two Brattle Square . Cambridge MA 02238-9105 617-547 5552 . FAX: 617-864-9405 California Office: 2397 Shattuck Avenue Suite 203 . Berkeley CA 94704-1567 . 510-843-1872 . FAX: 510-843-3785
February 22,1999 Page 2 of10 The NRC elected not to shut down the plants. According to an internal NRC document that we obtained via the Freedom ofInformation Act, the NRC's rationale in the Perry case was:
"The [2.206 petition review] board concluded that no urgent safety problems were uncovered that would wanant shutdown of the plant. The clad damage reported is insignificant and is allowable provided the Reactor Coolant Chemistry were within permissible Technical Specification (TS) limits as defined in TS 3.4.8. These limits are set to minimize radiological consequences of a postulated design bases accident and to meet appropriate acceptance criteria.
The petitioner does not allege that [ Perry Nuclear Power Plant] had operated outside the [ Technical Specifications]."' I fully agree that we never alleged that Perry, or River Bend for that matter, operated outside the Technical Specifications. Hence, the NRC's conclusion seems to have been based on things that we did ny say. But it is very disturbing that the NRC failed to address things that we d_l,,,d,, i say. In our River Bend petition, we specifically referenced ten separate sections within the plant's Updated Final Safety Analysis Report (UFSAR) that we felt are being violated by operation with failed fuel. In the attachment to our Perry petition, we referenced more than a dozen separate UFSAR sections, but we did not reference a single Technical Specification section. The documentation that we obtained on the NRC's evaluation does not contain a single word about these UFSAR violations. The Technical Specifications are part of the operating license. They define conditions that must be satisfied for the plant to operate. Plant owners can change the Technical Specifications only after formal review and approval by the NRC. The Updated Final Safety Analysis Report (UFSAR) describes the plant's design features and the analyses performed of the plant's response to postulated accidents. It also defines conditions that must be satisfied for the plant to operate. The UFSAR is the primary document reviewed by the NRC in reaching the decision to grant an operating license. The NRC apparently failed to even look at this key document when it evaluated our petitions. We feel it is wrong to evaluate this concern solely on the bases ot' Technical Specification compliance. More importantly, the NRC also knows it is wrong. I call your attention to a notice sent by the NRC to all plant owners in March 1996: The U.S. Nuclear Regulatory Commission (NRC) is issuing this informttion notice to alert addressees to instances of reactor operation that may not conform to the licensing basis. It is expected that recipients will review the infonnation for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.
]
On August 21,1995, the NRC received a petitior. . , JR 2.206 which was supplemented on August 28,1995, that requested NRC to shut down Millstone Unit I and take enforcement action based upon alleged violations oflicensed activities related to operation of spent fuel pool cooling systems and refueling practices. Followup of the issues raised in the 2.206 petition, including the findings from investigations conducted by the Otiice of the Inspector General, found that certain activities at Millstone Unit I may have been conducted in violation oflicense requirements and that refueling activities may not have been conducted consistent with the Updated Final Safety Analysis Report (UFSAR).2 ' FOIA/PA 99-079 dated January 13,1999,"2.206 Petition Review Board Summary December 9,1999"
- NRC Information Notice No. 96-17," Reactor Operation Inconsistent with the Updated Final Safety Analysis Report," March 18,1996
February 22,1999 Page 3 of10 7 V When the NRC received the Millstone petitions in August 1995, they focused on Technical Specification ' issues even though the overwhelming majority of the concems involved design issues and analyses in the UFSAR. During the informal public hearing held in April 1996 for the Millstone 2.206 petition, Emie Hadley, the counsel for the petitioner, compared the plant's Technical Specifications to a driver's license and the Updated Final Safety Analysis Report to the Motor Vehicle Code. He contended that while the driver's license contains a few restrictions on the driver, such as the need to wear corrective lenses, it basically symbolizes the driver's commitment to follow the requirements in the Motor Vehicle Code in exchange for the privilege to operate a car. Hadley pointed out that when the driver fails to follow the requirements in the Motor Vehicle Code, the state can suspend the license, revoke it, or modify it such that the driver can operate a vehicle only to and from a job. Emie Hadley argued that Millstone's owners retained the privilege to operate the nuclear plant if and only if they conformed to the conditions within the Technical Specifications and the Updated Final Safety Analysis Report. Mr. Hadley has been an attomey for many years. He was not expressing his opinion, or his desire, or even his philosophy. He was stating his conviction of what the regulations required. History has clearly demonstrated that Mr. Hadley was absolutely right. The NRC was wrong in 1995 to focus exclusively on the Technical Specifications and they are still wrong to do so today. In response to both the River Bend and Perry petitions, the NRC indicated that continued reactor operation with leaking fuel was permissible because the radioactivity levels of the water being circulated through the reactor core remained below the Technical Specification limit. It is true that the reactor water radioactivity limits remain within the Technical Specification limits. It is also true that Annapolis is the [ml capitol of Maryland and there are 12 inches in a foot. None of these truths has any bearing on whether the River Bend and Perry plants can safely operate with leaking fuel. What does matter is whether reactor operation with failed fuel is consistent with the as-low-as-reasonably-achievable (ALARA) requirements for plant workers and with the input assumptions for accident analyses. We contended in our petitions, and still maintain, that there is a disconnect on both accounts. Before explaining these two disconnects, I will briefly provide some background on nuclear fuel design and past experiences with failed fuel. This overview is taken from our April 1998 report which was an attachment to each of the petitions. Nuclear plants are powered by fuel pellets roughly the size and shape of a large pencil eraser stacked within 12 to 14 feet long metal tubes sealed at each end with welded metal caps.' A simplified drawing of a fuel tube is shown in Figure 1 of our report. The fuel tubes are also called the fuel cladding. Fuel tubes are like the gas tank in a car - when the tank is breached, highly volatile gasoline spills out to threaten the safety ofits passengers and innocent bystanders. When fuel tubes are breached, he -ly radioactive material spills out to threaten the safety of plant workers and the public. All operating US nuclear power plants use fuel assemblies containing square arrays of fuel tubes. A typical fuel assembly is illustrated in Figure 2 of our report. According to the NRC, the fuel design bases ensure that "the fuel is not damaged as a result of normal operation and anticipated operational occurrences."' It does not say that the fuel will suffer only minor damage. It says that the fuel will noto be damaged during normal operation. Thus, the fuel design bases includes the explicit requirement for the fuel tubes to remain intact during normal operation. (\ ! v
' Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.3.2.1," Fuel Rod Mechanical Design," and General Electric Company, " Licensing Topical Report / General Electric Standard Application for Reactor Fuel," NEDO-24011-A-4, January 1982.
- Nuclear Regulatory Commission, NUREG-0800, Standard Review Plan, Section 4.2, Fuel System Design.
l
.1 February 22,1999 !
Page 4 of10 The splitting, or fissioning, of uranium atoms inside the fuel tubes releases energy that heats water - 0 atomic energy that powers the plant. Byproducts of the fission process include radioactive gases and solids. These radioactive materials emit gamma rays along with alpha and beta particles which can cause damage to the human body. The fuel tubes contain these radioactive materials. If the tubes break, radioactive materials leak into the water which cools them. This water is contained within the reactor vessel and the piping connected to it, which form a second barrier to contain the radioactive materials. If . the piping fails, contaminated water spills into the reactor containment building. The reactor vessel and ) its piping are located within a reactor containment building which forms a third barrier. Because the reactor containment building is not leak tight, it reduces, but does not eliminate, the possibility that radioactive material would escape. Figure 3 in our report shows a simplified drawing of these three ! barriers. The fuel tubes are the most important of the three barriers. If the fuel tubes remain intact, the other two barriers can completely fail and the public will still be protected. The intact fuel tubes contain the radioactive gases and solids and prevent them from being released to the atmosphere. The pubile cannot be harmed from a nuclear plant accident in which the fuel tubes remain intact. But the River Bend ; and Perry plants are operating with this vital barrier already broken. j { Leaking fuel tubes are detected by increased radioactivity levels in the reactor vessel's liquid and gaseous releases.' Not surprisingly, the radioactivity levels rise significantly when fuel tubes break. The l causes and precise locations of fuel tube failures cannot be determined until the plant is shut down and i the leaking fuel tubes examined. A few years ago, the owner of the Point Beach Nuclear Plant in Wisconsin reported a significant event in O which "The fuel cladding was failed to the extent that fuel pellets could be seen through the hole in the clad. However, no pellets escaped from the rod."'Ihe fuel tube's failure was detected when the radioactivity levels of the reactor water rose to a level that was "10 percent of that allowed by [the Technical Specifications]."' In other words, the plant's Technical Specifications would have allowed it to ! remain running with up to nine other similarly damaged fuel tubes. This event suggests that the ! restrictions on reactor water radioactivity levels are too high to prevent operation with gaping holes in l fuel tubes. l At the Palisades plant in Michigan, three portions of a broken fuel tube were discovered in different parts of the reactor. One segment, nearly 7% feet long, was missing about one-third ofits fuel pellets. A second segment,4K feet long, and a third segment,1 % feet long, appeared to contain all their fuel pellets.' This event is disturbing because it highlights how fragile fuel tubes can become during nonnal operation. At Palisades this damaged fuel tube literally fell apart as it was being removed from the reactor core and many ofits fuel pellets were lost. I'll return to the first of the two disconnects. This one involves radiation exposures to plant workers. l Nuclear plant owners are required by federal regulations to keep the release of radioactive materials "as low as reasonably achievable"(ALARA).' According to the NRC,"a plant operating with 0.125 percent 8 Entergy Operations, River Bend Station Updated Final Safety Analysis Report, Section 4.2.4.2,"Online Fuel System Monitoring," and Section 11.5.2.2.1," Main Steam Line Radiation Monitoring System."
- Wisconsin Electric Power Company, Licensee Event Report No. 85-002-01, " Failed Fuel Rod in Assembly H14, Point Beach Nuclear Plant Unit 1," May 19,1986.
' United States Nuclear Regulatory Commission, Information Notice 93-82, "Recent Fuel And Core Performance Problems in Operating Reactors," October 12,1993.
- Title 10 of the Code of Federal Regulations. Sections 50.34a," Design objectives for equipment to control releases
l Februaly 22,1999 Page 5 of 10
'j pin-hole fuel cladding defects showed a general five-fold increase in whole-body radiation exposure rates in sorne areas of the plant when compared to a sister plant with high-integrity fuel (<0.01 percent leakers). Around certain plant systems the degraded fuel may elevate radiation exposure rates even more."' The " sister plants" were virtually identical because they were built at the same time by the same i owner on the same site. The significant variation in radiation exposure rates is ny due to thicker concrete or other design differences - it is due to the damaged fuel tubes. This NRC evidence is troubling because it shows a significantly increased risk to nuclear plant workers at a facility operating withjust 0.125 percent fuel tube failures. Many plants consider it permissible to operate with eight times as many fuel tube failures (up to 1.0% of the fuel tubes damaged).
Holes and cracks in fuel tubes release radioactive materials into the reactor water. The water carries them , to all parts of the plant, contaminating equipment throughout the facility. Workers conducting equipment inspections and maintenance receive higher radiation exposures. Indeed, some plant workers have received radiation doses far greater than allowed by federal regulations.' At plants like River Bend and Perry, this means that it might take 10 workers to do a job that would normally be done by 5 workeis so that no individual worker receives an excessive radiation dose. But the work force receives more radiation exposure than if the damaged fuel was removed. According to Section 12.1.2.1, " General Design Considerations for ALARA Exposures," of the River I Bend UFSAR:
"The general design considerations and methods to be employed to maintain in-plant radiation exposure ALARA have two objectives:
) (1) Minimizing the amount of time plant personnel spend in radiation areas, (2) Minimizing radiation levels in routinely occupied plant areas and in the vicinity I of plant equipment expected to require the attention of plant personnel."
It is a well-documented fact that plant operation with damaged fuel tubes significantly increases ] personnel exposures. Federal regulations require nuclear plant owners to keep the release of radioactive materials as low as reasonably achievable. Therefore, it seems both an illegal activity and a serious health hazard for nuclear plants to continue operating with known fuel damage. There are precedents for taking this action. Just last year, the owner of the Limerick nuclear plant outside q Philadelphia elected to shut that plant down for about a week to remove leaking fuel bundles. That utility I is committed to operating nuclear plants with radiation levels ALARA. I'll turn now to the disconnect between River Bend and Perry operating with failed fuel and the design l bases upon which the NRC issued them an operating license. First, let me explain our understanding of the relationship between normal plant operating conditions and the associated design bases using an example. River Bend and Perry have an ASME Code pressure limit of 1375 psig for the reactor vessel. That doesn't mean that the plants can routinely operate with reactor pressure up to, but not exceeding, that limit. The plants must be able to accommodate the largest pressure spike that may occur from of radioactive material in effluents - nuclear power reactors," and 50.36, " Technical specifications," and Title 10 of the Code of Federal Regulations, Part 50, Appendix I," Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion "As Low As Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents." I m)
' United States Nuclear Regulatory Commission, Info;mation Notice No. 87-39, " Control Of Hot Particle Contamination At Nuclear plants," August 21,1987.
' United States Nuclear Regulatory Commission, Information Notice No. 87-39," Control Of Hot Particle Contamination At Nuclear plants," August 21,1987.
1 1 i February 22,1999 i Page 6 of 10 ! postulated transients. Section 15B.5.2.2, "Overpressurization Protection," of Perry's UFSAR explains - l that concept:
)I "The overpressure protection system must accommodate the most severe pressurization transient i so that the ASME code limit of 1375 psig is not exceeded. The Main Steam Isolation Valve l (MSIV) closure with secondary scram (tlux scram) has been determined to be the most limiting j event for overpressure protection."
{ What this means in English is that when the valves in the pipes that carry the steam from the reactor vessel to the turbine close, the pressure inside the reactor vessel goes up.
. \
According to Table 15.0-1, " Input Parameters and initial Conditions for Transients," in Perry's UFSAR, j the input parameters for this analysis essume the reactor vessel dome pressure is 1045 psig. UFSAR Table 15B.5.2-3, "Overpressurization Protection Analysis Results," reports a peak reactor vessel bottom pressure of 1289 psig for the postulated transient. The pressure numbers are not important to my point. The process, however, is very important. It features three elements: initial condition (1045 psig), effect of change (about 250 psig), and an end point (1289 ) psig) which demonstrate the safety margin to the design limit (1375 psig). These three elements form the link between nonnal plant operat ion and accident analyses. As long as the plant operates within th-bounds established by the initial conditions, the postulated accident or transient can occur with l reasonable assurance that the end point will have at least the minimum safety margins. i Containment temperature is another example. Page 6.2-3 of the River Bend UFSAR states that the containment design temperature is 185eF. UFSAR Table 6.2-3, " Conditions for Containment Response h Analyses" states that the accident analyses assume the containment air temperature at the beginning of the accident is 90oF and that it rises to 141oF following the accident. Again, the three elements are present: initial condition (90oF), effect of change (51oF), and end point (141oF) which provides safety margin to the design limit (185oF). So, the link between accident analyses and normal reactor operation allows plants to startup, change l power, and run as long as they remain within the initial conditions assumed in the analyses. As we read I the River Bend and Peny UFSARs, it appears that this link cannot be made unless the plants operate with zero fuel failures. Commissioner Diaz stated during a recent public meeting that zero defects is an unrealistic standard. I wholeheartedly agree with him. However, the burden is on the plant owners to define, and on the NRC to approve, a more realistic standard. The burden is not, repeat not, on the public to accept the NRC Ignoring violations of federal safety regulations. l Let's take a closer look for the missing link between safety analyses and normal reactor operation with respect to damaged fuel. Consider the "Rechenlation Flow Control Failure with Increasing Flow" event as described in Peny UFSAR Section 15.4.5:
" Failure of the master controller of neutron flux controller can cause an increase in the core coolant flow rate. Failure within a loop's flow controller can also cause an increase in core coolant flow rate."
In boiling water reactors like those at River Bend and Perry, increasing the amount of flow through the reactor core causes the power of the reactor to rise. Per Section 15.4.5.5, " Radiological Consequences, " g of the Perry UFSAR, "An evaluation of the radiological consequences was not made for this event, since no radioactive material is released from the fuel." Now, this statement cannot be true unless the initial conditions assume that there is no leaking fuel. If damaged fuelis present, radioactive material will
E y February 22,1999 h
. Page 7 of10 o
be released from the fuel. In fact, more radioactive material will be released due to this event because the release rate is primarily dependent on power level and the reactor power level will increase. And that's not all. Table 15.0-3, " Summary of Accidents," of the Perry UFSAR lists the number of failed fuel tubes from the safety analyses performed for various postulated accidents: Failed FuelTubes Rod Drop Accident <770 Steam System Pipe Break Outside Containment None Feedwater Line Break None LOCA Within RCPB . None Once again, these results would appear flawed unless the reactor is operating with no damaged fuel when the accident begins. In the control rod drop accident analysis, fewer than 770 fuel tubes were calculated to fail as a result of that event. But are plant workers and the general public still protected if the accident
^ occurs when there are already 3, or 5, or 10, a dozen, or a hundred leaking tubes? Will there be adequate protection if nearly 770 fuel tube failures are added to numerous pre-existing leaking tubes? I don't know. The plain truth is that no one knows.
Perry UFSAR Section 15.4.2.1.1, " Identification of Causes," describes the control rod withdrawal error event as:
"The rod withdrawal error (RWE) transient results from a procedural error by the operator in which a single control rod or a gang of control rods is withdrawn continuously until the rod withdrawal limiter (RWL) function of the rod control and information system (RC&IS) blocks further withdrawal."
In boiling water reactors like those at River Bend and Perry, withdrawing control rods uncovers portions of fuel tubes causing their power output to increase significantly. This local power increase effect may or may not cause the overall reactor core power level to increase. But the power level of the uncovered fuel tubes can increase by a factor of ten. Remember that power level is the primary factor controlling how much radioactivity escapes through holes or cracks in fuel tubes. Raising the local power level by a j factor of ten means that much more radioactivity will be released. How do the safety analyses account for this fact? Section 15.4.2.5, " Radiological Consequences," of the Perry UFSAR states: "An evaluation of the radiological consequences was not made for this event, since no radioactive material is released from the fuel." Section 15.4.2.5," Radiological Consequences," of the j River Bend UFSAR contains this statement word for word. They are both correct, but only when this ' event initiates with zero damaged fuel in the reactor core. When there is leaking fuel, the results for the safety analyses of this event are not applicable. Federal regulations do not allow nuclear plants to operate with non-applicable safety analyses.The NRC might, but the regulations do not. But don't take my word for it. Look at Table 15A.2-1,
" Unacceptable Consequences Criteria Plant Event Category: Nonnal Operation," in the River Bend UFSAR. According to River Bend's UFSAR, an unacceptable consequence during normal operation is:
" Existence of a plant condition not considered by plant safety analyses." As we just reviewed, River 7 Bend's safety analysis for the rod withdrawal error event consider the fuel tubes to be completely (O intact before, during, and after the event. The plant is currently not in a condition covered by its safety analysis. Neither is Perry.
i February 22,1999 Page 8 of10 , There are other unanalyzed consequences of having an accident with pre-existing leaking fuel. For O example, the fuel tubes are filled with helium prior to being sealed. Helium is used because ofits high i thermal conductivity." The leakage of helium through holes and cracks in the fuel tubes may slow down l the transfer of heat from the fuel pellets to the water. When this heat cannot be dissipated as quickly as { assumed, the fuel temperature will increase and may reach the point at which it begins to melt. The leakage of he!iitm from a fuel tube may reduce heat transfer rates, thus potentially increasing the chances
)
that the fuel will be seriously damaged during a loss-of-coolant accident. . How much margin is currently available. According to Table 15B.6.3-1 of the Perry UFSAR: l 1
"The maximum fuel peak clad temperature for the GE10 fuel is 2149eF plus the 6F l adders / penalties resulting in a total of 2155aF."
"The maximum fuel peak clad temperature for the Cycle 7 gel 1 fuel is 2184oF plus the 6*F )
adders / penalties resulting in a total of 2190oF."
"The maximum fuel peak clad temperature for the Cycle 7 GE12 fuel is 2181oF plus the 6'F adders / penalties tesulting in a total of 2187aF." l The maximum fuel peak clad temperature allowed by federal regulations is 2200aF. Thus, the margin at Perry is 10 to 45 degrees. Will the loss of helium gas reduce that margin by a degree, ten degrees, twenty degrees, or fifly degrees? I don't know. From our research, it appears that the NRC and the plant owners don't know either. I What do we know about reactors that operate with leaking fuel? We know that holes or cracks in the fuel O'
tube allow radioactivity and helium gas to leak og. They also allow water to leak in. The high temperature produced by an operating reactor core dissociates the water into hydrogen and oxygen gases. The hydrogen gas interacts with the metal fuel tube to form blisters. The blisters embrittle the metal, leading to perforations." Thus, water leaking into a fuel tube may increase the probability that it suffers this type of damage. In fact, failure propagation due to this cause has already been identified. In 1993, a fuel tube at the Perry plant experienced a crack measuring 20 inches long, or nearly 13% of the fuel tube's length, caused by this water intrusion sequence." So, it is known that small holes and cracks can propagate during normal plant operation into rather large cracks. When you release the inlet of a balloon, it moves. Air rushes from the inlet in one direction, but the balloon travels in the other direction. It's a basic law of physics - for every action, there is an equal and opposite reaction. If one of the pipes connected to the reactor vessel breaks, a classic accident scenario, water and steam would rush out of the opening. The forces inside the reactor vessel are much larger than during normal operation. In addition, these forces are in different directions than during normal operation. What happens when metal tubes, weakened by holes and cracks, are exposed to larger forces from new directions? Recall the fragile fuel tube at the Palisades plant that fell apart when it was picked up. Does this evidence suggest that damaged fuel tubes can withstand the forces inside the reactor vessel during an accident? At risk of sounding like a broken record or appearing stupid, I don't know. From our l L " Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.3.2.1," Fuel Rod Mechanical Design." " Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 1 3.7.2.1," Burnable Poison Rod Design Evaluation." l " United States Nuclear Regulatory Commission, Information Notice 93-82,"Recent Fuel And Core Performance Problems In Operating Reactors," October 12,1993. i i I
February 22,1999 Page 9 of10 b,, research, it appears that the NRC and the plant owners don't know either. In responding to our petitions, the NRC indicated that the plants did not need to be immediately shut down because they were within the Technical Specifications' limit on radioactivity levels of the water flowing through the reactor core. As I have stated, that is true. It is also irrelevant because there is no link between a plant operating just below that Technical Specification limit and the accident analyses. The Technical Specification limit corresponds roughly to one percent of the fuel tubes being damaged. Some of the accident analyses assume that the plant is initially operating with one percent of the fuel tubes damaged, but most of these analyses assume none of the fuel tubes are damaged when the accident begins. The analyses that assume no failures are clearly not bounding when the plant operates with failed fuel tubes. However, even the accident analyses that assume that one percent of the fuel tubes are damaged when the accident starts are also not bounding. As I have covered, pre-existing fuel tube failures could propagate as a dkeet result of an accident. Therefore, compliance with the Technical Specification limit does not mean that workers and the public will be protected from excessive radiation exposure in event of an accident. Conformance with the design bases requirements of the UFSAR is needed to ensure adequate protection. Plant operation with damaged fuel is not properly addressed in the River Bend and Perry UFSARs. There's not a single word, picture, chart, or table describing what happens when the helium gas leaks out through a hole or crack or what happens when water leaks in. Nothing. In the report we submitted to the l NRC last April, I documented a safety evaluation I performed per 10 CFR 50.59. This federal regulation controls whether plant owners can change how they operate their plants without prior NRC approval. Prior to joining UCS, I prepared and reviewed thousands of 50.59 safety evaluations. The'50.59 safety
) evaluation I performed for plants operating with damaged fuel clearly established that NRC approval is required. Neither River Bend nor Perry have sought, or be given, permission by the NRC.That's a violation of federal regulations that were created to protect public health and safety.
Although I have not yet heard it yet, the NRC and/or the plant owners may argue that the results from analysis performed for the classic loss ofcoolant accident sequence bounds all of the other accident consequences. If made, that argument would be fallacious. It is true that if that accident were to occur, the presence of pre-existing leaking fuel will not make a discernable difference in the amount of radioactivity released to the atmosphere. For example, a dozen or two damaged fuel bundles at Three Mile Island would not changed the millions of curies of radioactivity released when the reactor core melted down. But, there are other accident sequences that are not bound by this classic analysis. At many plants, the maximum radiological threat to the control room operators comes from either the control rod drop accident or the break of a steam line outside the containment building. The presence of a dozen or two leaking fuel bundles when these accidents occurs could mean that control room operators receive far more radiation than permitted by federal regulations. From published news accounts, the NRC and utility representatives have stated that River Bend and Perry are not the only plants to operate with damaged fuel - that plants have been operating with damaged fuel for many years. Their implication is that this history demonstrates that such operation is safe. That's simply not true. It could be luck, and not safety margins. There hasn't been an accident at a plant with damaged fuel. So, experience does not demonstrate that plants operating with damaged fuel is safe. And the safety analyses do n_ot o demonstrate that plants operating with damaged fuel is safe. The
'. obvious question, to us, is therefore why are River Bend and Perry operating with damaged fuel?
O. V Earlier, I compared the NRC actions on our petitions to their actions on the Millstone petitions. In that case also, the NRC and the plant owner argued that what Millstone was doing - offloading the entire reactor core to the spent fuel pool every refueling outage s okay because everyone was doing it. Atler further review, it was learned that many of the plar. .nat were doing it should not have been doing
February 22,1999 Page 10 of 10 it because they had not taken the proper precautions. Had an accident occurred, those plants may not have been able to prevent radiation releases that impacted public health. They were lucky. Luck is not an acceptable standard for public protection. In our petitions, we asked the NRC to order the River Bend and Perry plants to be immediately shut down. We did not ask the NRC to take away the keys. Instead, we asked the NRC to prevent the plants from restarting until the damaged fuel had been replaced by non-leaking fuel or until the plant owners had performed safety analyses which demonstrated workers and the public would be protected if an accident were to occur with the pre-existing leaking fuel. That's what the regulations require. We simply asked the NRC to stop being a spectator and start being a regulator. The situation at River Bend and Perry can be compared to health insurance. When an insurance company finds out that one ofits customers had a pre-existing health problem that was not accurately reported on the medical survey, it will either raise the premium or cancel the policy. It will take this action because the risk factor is higher with the pre-existing condition. The difference between this example and the situation at River Bend and Perry is that the insurance company will not overlook the matter if the customer correctly spelled all of the words on the medical survey. The insurance company, unlike the NRC, focuses on substance. If the NRC carefully reviews the facts in this matter, we are confident that they will take the necessary actions to protect public health and safety, If the NRC disagrees with our contentions, we would ask that they document how each one of the many UFSAR discrepancies that we have identified in our petitions and during this presentation are satisfied. We will not accept an NRC denial of our petition based exclusively on the fact that the plants comply with their Technical Specifications. The 2.206 petition process does not currently have an effective appeal process. So our appeal, like in the Millstone case, h will be to the media, to elected officials, and to the public. I have described several safety questions that must be answered before anyone can truthfully say that nuclear plants operating with damaged fuel do not pose a threat to workers and the general public. At this point, no one knows. Ignorance may be bliss, but bliss is not safe. What is really needed is a bliss reduction program. 1 O 1
[ (. sammmmm mmmm mmmmmmes O' UNION OF CONCERNED SCIENTISTS November 9,1998 Dr. William Travers Executive Director for Operations United States Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
PETITION PURSUANT TO 10 CFR 2.206, PERRY NUCLEAR POWER PLANT
Dear Dr. Travers:
The Union of Concemed Scientists submits this petition pursuant to 10 CFR 2.206 requesting that the Perry Nuclear Power Plant in Ohio be immediately shut down and its operating license suspended or modified until such time that the facility's design and licensing bases are properly updated to permit operation with failed fuel assemblies or until all failed fuel assemblies are removed from the reactor core. m
Background
On April 2,1998, UCS provided the Nuclear Regulatory Commission with a copy of our repon titled
" Potential Nuclear Safety Hazard / Reactor Operation with Failed Fuel Cladding." We concluded:
UCS considers nuclear plants operating with fuel cladding failures to be potentially unsafe and to be violating federal regulations. The NRC's Weekly Information Report for the week ending October 30,1998, stated: On September 2,1998, the [ Perry Nuclear Power Plant] licensee detected an increase in the long lived isotope Xe 133 in the offgas pretreatment steam, indicating the existence of a pinhole leak in a fuel rod.
"Ihe GE Fuel Performance Manager was consulted and concurred with the conclusion tbst a pinhole leak existed.
Subsequent examinations led the licensee to conclude that the leak came from a twice-bumed fuel bundle. The leak was suppressed by repositioning of the control rods. In addition, the licensee implemented administrative procedures to limit power changes to less than or equal to one percent per hour to prevent further degradation to the fuel pin. On October 28,1998, the licensee identified a second increase in noble gas thus indicating a second pinhole leak in a fuel rod. The licensee intends to take actions to identify and suppress this second leak
~
over the upcoming weekend. O V Washington Office: M16 P Street NW Suite 310 . Washington DC 200361495 e 202 332 0900 . FAX: 202 332 0905 i Cambridge Headquarters: Two Bratue Square . Cambridge MA 02238-9105 617 547 5552 . FAX: 617-864-9405 Califomia Office: 2397 Shattuck Avenue Suite 203 . Berkeley CA 947041567 510-8431872 . FAX: 510-843-3785 j i i
p
.c 5- November 9,1998 Page 2 of 3 No obvious cause of the failure can be determined at this time. De most common cause of fuel failures among BWRs is debris induced fretting. ne Perry facility has been operating at essentially full power since the fall 1997 refueling outage. De next refueling outage is scheduled for April 1999.
As detailed in UCS's April 1998 report on reactor operation with failed fuel cladding, it has not been demonstrated that the effects from design bases transients and accidents (i.e., hydrodynamic loads, fuel enthalpy changes, etc.) prevent pre-existing fuel failures from propagating. He available information for the Perry plant suggests that either the original fuel failure is propagating and/or there is a common mode failure mechanism degrading cladding integrity. it is therefore possible that significantly more radioactive material will be released to the reactor coolant system during a transient or accident than that experienced during steady state operation. %us, the existing design bases accident analyses for Perry do n_o3 bound its current operation with known fuel cladding failures. In addition to operating with non bounding design bases accident analyses, it appears that the Perry licensee is also violating its licensing basis for worker radiation protection under the as low as is reasonably achievable (ALARA) Progrmn. According to NRC Information Notice No. 87 39," Control of Hot Particle Contamination at Nuclear Plants:" A plant operating with 0.125 percent pin-hole fuel cladding defects showed a five fold increase in whole-body radiation exposure rates in some areas of the plant when compared to a sister plant with high-integrity fuel (<0.01 percent leakers). Around certain plant systems the degraded fuel may elevate radiation exposure rates even more. Industry experience demonstrated that reactor operation with failed fuel cladding increased radiation exposures for O plant workers. The Perry licensee informed the NRC about potential fuel cladding failures. It could shut down the facility and remove the failed fuel assemblics from the reactor core. Instead, it continues to operate the facility with higher radiation levels that are known to provide greater risk to plant workers. Since it appears that operation with one or more failed fuel assemblies is not permitted by its design and licensing bases, Perry must be immediately shut down. De facility must remain shut down until: O The Perry licensee removes the failed fuel assemblies from the reactor core.
-OR-0 The Perry licensee properly updates the plant's design and licensing bases to permit the plant to operate with known fuel damage.
Basis for Requested Action ~ l UCS is a non profit, public-interest organization with sponsors across the United States, including Ohio. UCS monitors performance at nuclear power plants in the United States against safety regulations promulgated by the NRC to protect the public and plant workers. When real or potential erosion of ' mandated safety margins is detected, as is currently indicated at this time at Perry, UCS engages the NRC, the media, and other authorities to resolve the safety concerns. Requested Actions UCS petitions the NRC to require the, Perry Nuclear Pc rer Plant to be immediately shut down snd that the facility remain shut down until all of the failed fuel assemblies are removed from the reactor core. Alternatively, the plant could be restarted after its design and licensing bases were properly updated to O reflect continued operation with failed fuel assemblies. I
.- November 9,1998 Page 3 of 3 UCS respectfully requests a hearing on this petition to preser,t new information on reactor operation with failed fuel assembhes. *Ihis new information will include, but is not limited to, a discussion of the April 1998 UCS report and the plant-specific information regarding Perry. While our concems apply to Perry, we respectfully request that this hearing be held in the DC area since the issue affects a3 operating nuclear power plants.
Sincerely, David A.Lochb um Nuclear Safety Engineer enclosure: " Potential Nuclear Safety Hazard / Reactor Operation with Failed Fuel Cladding," April 22,1998 O O .
r UNION OF Reactor Operation with Failed Fuel:- CONCERNED Bliss Reduction Program Needed SCIENTISTS Good aftemoon. My name is David Lochbaum. I have been the Nuclear Safety Engineer for the Union of Concemed Scientists (UCS) since October of 1996. Prior tojoining UCS, I worked in the nuclear industry for more than 17 years, primarily as a consultant. As a consultant, I worked on assignments at the Perry plant in 1995 and 1996. For one of these assignments, I developed a lesson plan on design and licensing bases requirements and presented it to managers, supervisors, and staffin the Design Engineering department. I was a reactor engineer for a total of nearly 8 years at the Hatch, Browns Ferry, Grand Gulf, and Hope Creek nuclear plants - all boiling water reactors similar in design to the River Bend and Perry plants. Among other reactor engineering duties, I was responsible for the fuel integrity mocitoring program and updating the accident analyses for each operating cycle. I am here today because UCS submitted two petitions to the NRC last year. Last spring, we provided the NRC with our technical report documenting our concerns with nuclear plants that operate with damaged
. fuel. At that time, we were not aware of any plants operating with damaged fuel. But plants had operated with damaged fuel in the past and were likely to operate with damaged fuel in the future. Five months later, we teamed that the River Bend plant was operating with damaged fuel leaking radiation. Other than a letter from the NRC acknowledging receipt of our technical report, we had not heard from the agency.
We still don't know if they agreed with, disagreed with, or even understood the concems. In an effort to prompt the NRC to take some action on our safety concerns, we submitted a pett jn asking that the River Bend plant be immediately shut down until the leaking fuel was removed or until the plant's owners had performed an analysis showing that it was safe to operate with the , leaking fuel. A few weeks later, we teamed that the Perry plant was also operating with failed fuel. So, {' we submitted a similar petition. It is our intention to continue filing petitions when plants operate with leaking fuel until the NRC addresses our concerns or until we run out of postage stamps. And we have plenty of stamps. l i The River Bend plant, located north of Baton Rcuge, Louisiana, was granted an operating license by the NRC on November 20,1985. The Perry plant, lo:ated northeast of Cleveland, Ohio, was issued an
]
l operating license about a year later on November 13,1986. In each case, the NRC issued the operating j license after a lengthy, deliberative process through which it concluded there was reasonable assurance l that two criteria were met: 1
- 1. That the facility's design met all applicable regulatory requirements. j
- 2. That the facility would be operated and maintained in accordance with all applicable !
regulatory requirements. { By letters dated September 25,1998, and November 9,1998, UCS petitioned the NRC to require the immediate shut down of the River Bend and Perry p: ants because they continue to operate even though some of their nuclear fuel is leaking radioactivity - a condition that violates the terms af their license. As a result of this illegal activity, plant workers face greater risk from radiation overexposure during normal plant operation. In addition, both plant workers and the general pu'sile may face greater risk from radiation overexposure if an accident were to occur. I i Washington Office: 1616 P Street NW Suite 310 . Washington DC 20036-1495 e 202 332 0900 . FAX: 202 332 0905 Carnbridge Headquarters Two Brattle Square . Cambridge MA 02238-9105 617 547 5552 . FAX: 617-864 9405 California Office: 2397 Shattuck Avainue Suite 203 . Berkeley CA 94704 1567 . 510-843-1872 . FAX: 510-843-3785 ENCLOSURE 3
i February 22,1999 Page 2 of10 The NRC elected not to shut down the plants. According to an intemal NRC document that we obtained via the Freedom ofInformation Act, the NRC's rationale in the Perry case was:
"The [2.206 petition review] board concluded that no urgent safety problems were uncovered that would warrant shutdown of the plant. The clad damage reported is insignificant and is allowable provided the Reactor Coolant Chemistry were within permissible Technical Specification (TS) limits as defined in TS 3.4.8. These limits are set to minimize radiological consequences of a postulated design bases accident and to meet appropriate acceptance criteria.
The petitioner does not allege that [ Perry Nuclear Power Plant] had operated outside the [ Technical Specif' cations]."' I fully agree that we never alleged that Perry, or River Bend for that matter, operated outside the Technical Specifications. Hence, the NRC's conclusion seems to have been based on things that we did ny say. But it is very disturbing that the NRC failed to address things that we d, lid say. In our River Bend petition, we specifically referenced ten separate sections within the plant's Updated Final Safety Analysis Report (UFSAR) that we felt are being violated by operation with failed fuel. In the attachment to our Perry petition, we referenced more than a dozen separate UFSAR sections, but we did not reference a single Technical Specification section. The documentation that we obtained on the NRC's evaluation does not contain a single word about these UFSAR violations. The Technical Specifications are part of the operating license. They define conditions that must be satisfied for the plant to operate. Plant owners can change the Technical Specifications only after formal review and approval by the NRC. The Updated Final Safety Analysis Report (UFSAR) describes the plant's design features and the analyses performed of the plant's response to postulated accidents. It also defimes conditions that must be satisfied for the plant to operate. The UFSAR is the primary document reviewed by the NRC in reaching the decision to grant an operating license. The NRC apparently failed to even look at this key document when it evaluated our petitions. We feel it is wrong to evaluate this concern solely on the bases of Technical Specification compliance. More importantly, the NRC also knows it is wrong. I call your attention to a notice sent by the NRC to all plant owners in March 1996: The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to instances of reactor operation that may not conform to the licensing basis. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. On August 21,1995, the NRC received a petition under 10 CFR 2.206 which was supplemented on August 28,1995, that requested NRC to shut down Millstone Unit I and take enforcement action based upon alleged violations oflicensed activities related to operation of spent fuel pool cooling systems and refueling practices. Followup of the issues raised in the 2.206 petition, including the findings from investigations conducted by the Office of the Inspector General,
' found that certain activities at Millstone Unit I may have been conducted in violation oflicense requirements and that refueling activities may not have been conducted consistent with the Updated Final Safety Analysis Report (UFSAR).' ' FOIA/PA 99-079 dated January 13,1999,"2.206 Petition Review Board Summary December 9,1999" ' NRC Infonnation Notice No. 96-17 " Reactor Operation inconsistent with the Updated Final Safety Analysis Report," March 18,1996
February 22,1999 Page 3 of10 When the NRC received the Millstone petitions in August 1995, they focused on Technical Specification
- issues even though the overwhelming majority of the concerns involved design issues and analyses in the Ul'SAR. During the informal public hearing held in April 1996 for the Millstone 2.206 petition, Emie Hadley, the counsel for the petitioner, compared the plant's Technical Specifications to a driver's license and the Updated Final Safety Analysis Report to the Motor Vehicle Code. He contendeJ . hat while the driver's license contains a few restrictions on the driver, such as the need to wear correv.ve lenses, it basically symbolizes the driver's commitment to follow the requirements in the Motor Vehicle Code in - exchange for the privilep to operate a car. Hadley pointed out that wh;n the driver fails to follow the requirements in the Motor Vehicle Code, the state can suspend the license, revoke it, or modify it such that the driver can operate a vehicle only to and from a.iob.
Emie Hadley argued that Millstone's owners retainea tne privilege to operate the nuclear plant if and only if they conformed to the conditions within the Technical Specifications and the Updated Final Safety Analysis Report. Mr. Hadley has been an attorney for many years. He was not expressing his opinion, or his desire, or even his philosophy. He was stating his conviction of what the regulations required. History has clearly demonstrated that Mr. Hadley was absolutely right. The NRC was wrong in 1995 to focus exclusively on the Technical Specifications and they are still wrong to do so today. In response to both the River Bend and Perry petitions, the NRC indicated that continued reactor operation with leaking fuel was permissible because the radioactivity levels of the water being circulated through the reactor core remained below the Technical Specification limit. It is true that the reactor water radioactivity limits remain within the Technical Specifiention limits. It is also true that Annapolis is the capitol of Maryland and there are 12 inches in a foot. None of these truths has any bearing on whether i the River Bend and Perry p1rnts can safely operate with leaking fuel. What does matter is whether reactor operation with failed fuel is consistent with the as-low-as-reasonably-achievable (ALARA) ! requirements for plant workers and with the input assumptions for accident analyses. We contended in our petitions, and still maintain, that there is a disconnect on both accounts.
- Before explaining these two disconnects, I will briefly provide some background on nuclear fuel design and past experiences with failed fuel. This overview is taken from our April 1998 report which was an j attachment to each of the petitions. Nuclear plants are powered by fuel pellets roughly the size and shape of a large pencil eraser stacked within 12 to 14 feet long metal tubes scaled at each end with welded metal caps.' A simplified drawing of a fuel tube is shown in Figure 1 of our report. The fuel tubes are also called the fuel cladding. Fuel tubes are like the gas tank in a car - when the tank is breached, highly volatile gasoline spills out to threaten the safety ofits passengers and innocent bystanders. When fuel tubes are breached, highly radioactive material spills out to threaten the safety of plant workers and the public.
All operating US nuclear power plants use fuel assemblies containing square arrays of fuel tubes. A typical fuel assembly is illustrated in Figure 2 of our report. According to the NRC, the fuel design bases ensure that "the fuel is not damaged as a result of normal operation and anticipated operational occurrences. ' It does not say that the fuel will suffer only minor damage. It says that the fuel will not be damaged during normal operation. Thus, the fuel design bases includes the explicit requirement for the
' fuel tubes to remain intact during normal operation.
3 Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Mport, Section 3.3.2.1, " Fuel Rod Mechanical Design," and General Electric Company, " Licensing Topical Repon / Caneral Electric Standant Application for Reactor Fuel," NEDO.24011 A-4, January 1982.
- Nuclear Regulatory Commission, NUREG.0800, Standard Review Plan, Section 4.2, Fuel System Design.
I February 22,1999 Page 4 of10 i The splitting, e fissioning, of uranium atoms inside the fuel tubes releases energy that heats water - atomic energy that powers the plant. Byproducts of the fission process include radioactive gases and solids. These radioactive materials emi: gamma rays along with alpha and beta particles which can cause damage to the human body. The fuel tubes contain these radioactive materials. If the tubes break, radioactive materials leak into the water which cools them. This water is contained within the reactor vessel and the piping connected to it, which form a second barrier to contain the radioactive materials. If the piping fails, contaminated water spills into the reactor containment building. The reactor vessel and its piping are located within a reactor containment building which forms a third barrier. Because the reactor containment building is not leak tight, it reduces, but does not eliminate, the possibility that radioactive material would escape. Figure 3 in our report shows a simplified drawing of these three barriers. The fuct tubes are the most important of the three barriers. If the fuel tubes remain intact, the other two baniers can completely fail and the public will still be protected. The intact fuel tubes contain the radioactive gases and solids and prevent them from being teleased to the atmosphere. The public cannot be harmed from a nuclear plant accident in which the fuel tubes remain intact. But the River Bend and Perry plants are operating with this vital barrier already broken. Leaking fuel tubes are detected by increased radioactivity levels in the reactor vessel's liquid and 8 gaseous releases Not surprisingly, the radioactivity levels rise significantly when fuel tubes break. The causes and precise locations of fuel tube failures .:annot be determined until the plant is shut down and the leaking fuel tubes examined. A few years ago, the owner of the Point Beach Nuclear Plant in Wisconsin reported a significant event in which "The fuel cladding was failed to the extent that fuel pellets could be seen through the hole in the clad. However, no pellets escaped from the rod." The fuel tube's failure was detected when the radioactivity levels of the reactor water rose to a level that was "10 percent of that allowed by [the Technical Specifications]."'In other words, the plant's Technical Specifications would have allowed it to remain running with up to nine other similarly damaged fuel tubes. This event suggests that the restrictions on reactor water radioactivity levels are too high to prevent operation with gaping holes in j fuel tubes. { l At the Palisades plant in Michigan, three portions of a broken fuel tube were discovered in different parts of the reactor. One segment, nearly 5% feet long, was missing about one-third ofits fuel pellets. A second segment,4% feet long, and a third segment,1% feet long, appeared to contain all their fuel pellets.' This event is disturbing because it highlights how fragile fuel tubes can become during normal operation. At Palisades, this damaged fuel tube literally fell apart as it was being removed from the reactor core and trany ofits fuel pellets were lost. I'll r.:tum to the first of the two disconnects. This one involves radiation exposures to plant workers. Nuclear plant owners are required by federal regulations to keep the release of radioactive materials "as low as reasonably achievable"(ALARA).' According to the NRC, "a plant operating with 0.125 percent
' Entergy Operations, River Bend Station Updated Final Safety Analysis Report, Section 4.2.4.2, "Online Fuel
- System Monitoring." and Section 11.5.2.2.1," Main Steam Line Radiation Monitoring System."
' Wisconsin Electric 1 ower Company, Licensee En.h Report No. 85-002-01," Failed Fuel Rod in Assembly H14, Point Beach Nuclear Plant Unit 1," May 19,1986.
7 United States Nuclear Regulatory Commission, Information Notice 93-82, "Recent Fuel And Core Performance Problems In Operating Reactors," October 12,1993.
' Title 10 of the Code of Federal Regulations, Sections 50.344. " Design objectives for equipment to control releases i
Febmary 22,1999. Page 5 of10 pin-hole fuel cladding defects showed a general 'ive-fold increase in whole-body radiation exposure rates in some areas of the plant when compared to a sister plant with high-integrity fuel (<0.01 percent lenkers). Around certain plant systems the degraded fuel may elevate radiation exposure rates even more."' The " sister plants" were virtually identical because they were built at the same time by the same owner on the same site. The significant variation in radiation exposure rates is not due to thicker concrete or other design differences -it is due to the damaged fuel tubes. This NRC evidence is troubling because it shows a significawy increased risk to nuclear plant workers at a facility operating withjust 0.125 percent fuel tube <ailures. Many plants consider it permissible to operate with eight times as many fuel ! tube failures (up to 1.0% of the fuel tubes damaged). Holes and cracks in fuel tubes release radioactive materials into the reactor water. The water carries them to all parts of the plant, contaminating equipment throughout the facility. Workers conducting equipment inspections and maintenance receive higher radiation exposures. Indeed, some plant workers have received radiation doses far greater than allowed by federal regulations. At plants like River Bend and ' Perry, this means that it might take 10 workers to do a job that would normally be done by 5 workers so
- that no individual worker receives an excessive radiation dose. But the work force receives more radiation exposure t' m if the damaged fuel was removed.
According to Section 12.1.2.1, " General Design Considerations for. ALARA Exposures," of the River Bend UFSAR:
"The g neral design considerations and methods to be employed to maintain in-plant radiation exposure. ALARA have two' objectives:
(1) Minimizing the amount of time plant personnel spend in radiation areas, (2) MininCzing radiation levels in routinely occupied plant areas and in the vicinity of plant equipment expected to require the attention of plant personnel." It is a well-documented fact that plant operation with damaged fuel tubes significantly increases personnel exposures. Federal regulations require nuclear plant owners to keep the release of radioactive materials as low as reasonably achievable. Therefore, it seems both an ill+ gal activity and a serious health hazard for nuclear plants to continue operating with known fuel damage. There are precedents for taking this action. Just last year, the owner of the Limerick nuclear plant outside Philadelphia elected to shut that plant down for about a week to remove leaking fuel bundles. That utility is committed to operating nuclear plants with radiation levels ALARA. I'll turn now to the disconnect between River Bend and Perry operating with failed fuel and the design bases upon which the NRC issued them an operating license. First, let me explain our understanding of the relationship between normal plant operating conditions and the associated design bases using an example. River Bend and Perry have an ASME Code pressure limit of 1375 psig for the reactor vessel. That doesn't mean that the plants can routinely operate with reactor pressure up to, but not exceeding, that limit. The plants must be able to accommodate the largest pressure spike that may occur from of radioactive material in effluents - nuclear power reactors," and 50.36, "Te::hnical specifications," and Title 10 of the Code of Federal Regulations, Pan 50, Appendix I," Numerical Guides for Design Objectives anci Limiting Conditions for Operation to Meet the Criterion "As Low As Reasonably Achievable" for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents."
' United States Nuclear Regulatory Commission. Information Notice No. 87-39, " Control Of Hot Panicle Contammation At Nuclear plants," August 21,1987.
United States Nuclear Regulatory Commission, Information Notice No. 87-39, "Connol Of Hot Particle Contamination At Nuclear plants," August 21,1987.'
. _ _ _ . _ _ _ . _ _ 1
1 February 22,1999 Page 6 of10 ' postulated transients. Section 15B.S.2.2, "Overpressurization Protection," of Perry's UFSAR explains ! that concept: ;
"The overpressure protection system must accommodate the most severe pressurization transient i so that the ASME code limit of 1375 psig is not exceeded. The Main Steam Isolation Valve (MSIV) closure with secondary scram (flux scram) has been determined to be the most limiting event for overpressure protection."
What this means in English is that when the valves in the pipes that carry the steam from the reactor ; vessel to the turbine close, the pressure inside the reactor vessel goes up. ; According to Table 15.0-1, " Input Parameters and Initial Conditions for Transients," m Perr/s UFSAR, - i the input parameters for this analysis assume the reactor vessel dome pressure is 1045 psig. UFSAR Table 15B.S.2-3,"Overpressurization Protection Analysis Results," reports a peak reactor vessel bottom pressure of 1289 psig for the postu;ated transient. ' l He pmssure numbers are not important to my point. The process, however, is very important. It features three elements: initial condition (1045 psig), effect of change (about 250 psig), and an end point (1289 psig) which demonstrate the safety margin to the design limit (1375 psig). nese three elements form the link between normel plant operation and accident analyses. As long as the plant operates within the bounds established by the initial conditions, tne postulated accident or transient can occur with reasonable assurance that the end point will have at least the minimum safety margins. l Containment :emperature is another example. Page 6.2-3 of the River Bend UFSAR states that the containment design temperature is 185'F. UFSAR Table 6.2-3, " Conditions for Containment Response Analyses" states that the accident analyses issume the containment air temperature at the beginning of the accident is 90*F and that it rises to 141*F following the accident. Again, the three elements are present: initial condition (90*F), effect of change (51*F), and end point (141*F) which provides safety margin to the design limit (185'F). So, the link between accident analyses and normal reactor operation allows plants to startup, change power, and run as long as they remain within the initial conditions assumed in the analyses. As we read the River Bend and Perry UFSARs, it appears that this link cannot be made unless the plants operate with zero fu6 failures. Commissioner Diaz stated during a recent public meeting that zero defects is an unrealistic standard. I wholeheartedly agree with him. However, the burden is on the plant owners to define, and on the NRC to approve, a more realistic standard. The burden is not, repeat not, on the public to accept the NRC igacring violations cf federal safety regulations. Let's take a closer look for the missing link between safety analyses and normal reactor operation with respect to damaged fuel. Consider the " Recirculation Flow Coritrol Failure with Increasing Flow" event as described in Perry UFSAR Section 15.4.5:
" Failure of the master controller of neutron flux controller can cause an increase in the core coolant flow rate. Failure within a loop's flow controller can also cause an increase in core coolant flow rate."
In boiling water reactors like those at River Bend and Perry, increasing the amount of flow through the reactor core causes the power of the reactor to rise. Per Section 15.4.5.5, " Radiological Consequences, " of the Perry UFSAR,"An evaluation of the radiological consequence: xas not made for this event, since no radioactive material is released from the fuel." Now, thl: statement cannot be true unless the initial conditions assun e that there is no leaking fuel. If damaged fuel is present, radioactive material will
February 22,1999
. Page 7 of10 . be rek . x.: from the fuel. In fact, more radioactive material will be released due to this event becau:ie - the release rate is primarily dependent on power level and the reactor power level will increase.
And that's not all. Table 15.0-3, " Summary of Accidents," of the Perry UFSAR lists the number of failed rint tubes from the safety analyses performed for various postulated accidents: Failed FuelTubes Rod Drop Accident <770 Steam System Pipe Break Outside Containment None Feedwater Line Break None I LOCA Within RCPB . None Once again, these results would appear flawed unless the reactor is operating with no damaged fuel when I the accident begins. In the control rod drop accident analysis, fewer than 770 fuel tubes were calculated to fail as a result of that event. But we plant workers and the general public still protected if the accident occurs when there are already 3, or 5, or 10, a dozen, or a hundred leaking tubes? Will there be adequ te protection if nearly 770 fuel tube failures are added to numerous pre-existing leaking tubes? I don't know. The plain truth is that no one knows. Perry UFSAR Section 15.4.2.1.1, " Identification of Causes," describes the control rod withdrawal error event as:
"'Ihe rod withdrawal error (RWE) transient results from a procedural error by the operator in which a single control rod or a gang of control rods is wit:lrawn continuously until the rod 3 withdrawal limiter (RWL) function of the rod control and hformation system (RC&IS) blocks I further withdrawal."
In boiling water reactors like those at River Bend and Perry, withdrawing control rods uncovers portions of fuel tubes causing their power output to increase significantly. This local power increase effect may or may not cause the overall reactor core power level to increase. But the power level of the ancovered fuel tubes can increase by a factor of ten. Remember that power level is the primary factor controlling how much radioactivity escapes through holes or cracks in fuel tubes. Raising the local power level by a factor of ten means that much more radioactivity will be released. How do the safety analyses account for this fact? Section 15.4.2.5, " Radiological Consequences," of the Perry UFSAR states: "An evaluation of the radiological consequences was not made for this event, since no radioactive material is released from the fuel." Section 15.4.2.5, " Radiological Consequence's," of the River Bend UFSAR contains this statement word for word. They are both correct, but only when this event initiates with zero damaged fuel in the reactor core. When there is leaking fuel, the results for the safety analyses of this event are not applicable. Federal regulations do not allow nuclear plants to operate with non-applicable safety analyses. The NRC might, but the regulations do not. But don't take my word for it. Look at Table 15 A.2 1,
" Unacceptable Consequences Criteria Plant Event Category: Normal Operation," in the River Bend UFSAR. According to River Bend's UFSAR, an unacceptable consequence during normal operation is:
" Existence of a plant condition not considered by plant safety analyses." As we just reviewed, River Bend's safety analysis for the rod withdrawal error event consider the fuel tubes to be completely intact before, during, and after the event. The plant is currently not in a condition covered by its safety analysis. Neither is Perry.
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i February 22,1999 Page 8 of10 There are other unanalyzed consevenes of having an accident with pre-existing leaking fuel. For example, the fuel tubes are filled with helium prior to being sealed. Helium is used because ofits high thermal conductivity." ne leakage of helium through holes and cracks in the fuel tubes may slow down the transfer of heat from the fuel pellets to the water. When this heat cannot be dissipated as quickN as ! assumed, the fuel temperature will increase and may reach the point at which it begins to melt. The leakage of helium from a fuel tube may reduce heat transfer rates, thus potentially increasing the ch sces that the fuel will be seriously damaged during a loss-of-coolant accident. How much margin is currently available. According to Table 15B.6.3-1 of the Perry UFSAR:
"Tne maximum fuel peak clad temperature for the GE10 fuel is 2149'F plus the 6F adders / penalties resulting in a total of 2155'F." {
j "The maximum fuel peak clad temperature for the Cycle 7 gel 1 fuel is 2184*F plus the 6*F adders / penalties resulting in a total of 2190*F."
"The maximum fuel peak clad temperature for the Cycle 7 GE12 fuel is 2181*F plus the 6*F adders / penalties resulting in a total of 2187'F."
The maximum fuel peak clad temperature allowed by federal regulations is 2200*F. Thus, the margin i at Perry is 10 to 45 degrees. Will the loss of helium gas reduce that margin by a degree, ten degrees, i twenty degrees, or fifty degrees? I don't know. From our research, it appears that the NRC and the plant owners don't know either.
-What do we know about reactors that operate with leaking fuel? We know that holes or cracks in the fuel tube allow radioactivity and helium gas to leak o_ut. u They also allow water to leak ,in. The high temperature produced by an operating reactor core dissociates the water into hydrogen and oxygen gases.
The hydrogen gas interacts with the metal fuel tube to form blisters. The blisters embrittle the metal, leading to perforations." nus, water leaking into a fuel tube may incresse the probability that it suffers ! this type of damage. In fact, failure propagation due to this cause has already been identified. In 1993, a 1 fuel tube at the Peny plant experienced a crack measuring 20 inches long, or nearly 13% of the fuel tube's length, caused by tha water intrusion sequence." So, it is known that small holes and cracks can ; propagate during normal plant operation into rather large cracks. When you release the inlet of a balloon, it moves. Air rushes from the inlet in one direction, but the balloon travels in the other direction. It's a basic law of physics - for every action, there is an equal and opposite reaction. If one of the pipes connected to the reactor vessel breaks, a classic accident scenario, water and steam would rush out of the opening. The forces inside the reactor vessel are much larger than during normal operation. In addition, these forces are in different directions than during normal l operation. What happens when metal tubes, weakened by holes and cracks, are exposed to larger forces 1 from new directions? Recall the fragile fuel tube at the Palis, des plant that fell apast when it was picked up. Does this evidence suggest that damaged fuel tubes can withstand the forces inside the reactor vessel I during an accident? At ' risk of sounding like a broken record or appearing stupid, I don't know. From our
" Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.3.2.1, " Fuel Rod Mechanical Design." " Baltimore Gas & Electric Company, Calvert Cliffs Nuclear Plant Updated Final Safety Analysis Report, Section 3.7.2.1," Burnable Poison Rod Design Evaluation." " United States Nuclear Regulatory Commission, Information Notice 93 82,"Recent Fuel And Core Performance Problems In Operating Reactors," October 12,1993.
y February 22,1999 Page 9 of10 l research, it appears that the NRC and the plant owners don't know either. s In responding to our petitions, the NRC indicated that the plants did not need to be immediately shut l
. down because they were within the Technical Specifications' limit on radioactivity levels of the water {
flowing through the reactor core. As I have stated, that is true. It is also irrelevant because there is no link ~
' between a plant operating just below that Technical Specification limit and the accident analyses. The ,
Technical Specification limit corresponds rcugnly te one percent of the fuel tubes being damaged. Some
~of the accident analyses assume that the plant is initial;y operating with one percent of the fuel tubes damaged, but most of these analyses assume none of the fuel tubes are damaged when the accident 1 begins. The analyses that assume no failures are A .ly not bounding when the plant operates with failed i fuel tubes. However, even the accident analyses tant assume that one percent of the fuel tubes are
~ damaged when the accident starts are also not bounding. As I have covered, pre-existing fuel tube failures could propagate as a direct result of an accident. Therefore, compliance with the Techsleal Specincation limit does ng mean that workers and the public will be protected from excessive l
radiation exposure in event of an accident. Conformance with the design bases requirements of the UFSAR is needed to ensure adequate protection. Plant operation with damaged fuel is not properly addressed in the River Bend and Perry UFSARs. l There's not a single word, picture, chart, or table describing what happens when the helium gas leaks out through a hole or crack or what happens when water leaks in. Nothing. In the report we submitted to the ; NRC last April, I documented a safety evaluation I performed per 10 CFR 50.59. This federal regulation controls whether plant owners can change how they operate their plants without prior NRC approval. Prior to joining UCS, I prepared and reviewed thousands of 50.59 safety evaluations. The'50.59 safety evaluation I performed for plants operating with damaged fuel clearly established that NRC approval is required. Neither River Bend nor Perry' have sought, or be given, permission by the NRC. That's a violation of federal regulations that were created to protect public health and safety. Although I have not yet heard it yet, the NRC and/or the plant owners may argue that the results from analysis performed for the classic loss of coolant accident sequence bounds all of the other accident consequences. If made, that argument would be fallacious. It is true that if that accident were to occur, the presence of pre-existing leaking fuel will not make a discernable difference in the amount of radioactivity released to the atmosphere. For example, a dozen or two damaged fuel bundles at Three Mile Island would not changed the millions of curies of radioactivity released when the reactor core melted down. But, there are other accident sequences that are not bound by this classic analysis. At many plants, the maximum radiological threat to the control room operators comes from either the control rod drop accident or the break of a steam line outside the containment building. The presence of a dozen or two leaking fuel bundles when these accidents occurs could mean that control room operators receive far 1 more radiation than permitted by federal regulations. j From published news accounts, the NRC and utility representatives have stated that River Bend and Perry are not the only plants to operate with damaged fuel- that plants have been operating with i damaged fuel for many years. Their implication is that this history demonstrates that such operation is , safe. That's simply not true. It could be luck, and not safety margins. There hasn't been an accident at a I plant with damaged fuel. So, experience does not demonstrate that plants operating with damaged fuel is i safe. And the safety ani. lyses do no_t demonstrate that plants operating with damaged fuel is safe. The l obvious question, to us, is therefore why are River Bend and Perry operating with damaged fuel? ; Earlier, I compared the NRC actions on our petitions to their actions on the Millstone petitions. In that l case also, the NRC and the plant owner argued that what Millstone was doing- offloading the entire reactor core to the spent fuel pool every refueling outage - was okay because everyone was doing it. After further review, it was teamed that many of the plants that were doing it should not have been doing
February 22,1999 Page 10 of10 l it because they had not taken the proper precautions. Had an accident occurred, those plants may not have been able to prevent radiation releases that impacted public health. They were lucky, Luck is og an acceptable standard for public protection. , i In our petitions, we asked the NRC to order the River Bend and Perry plants to be immediately shut down. We did not ask the NRC to take away the keys. Instead, we asked the NRC to prevent the plants from restarting until the damaged fuel had been replaced by non-leaking fuel or until the plant owners - had performed safety analyses which demonstrated workers and the public would be protected if an accident were to occur with the pre-existing leaking fuel. That's what the regulations require. We simply asked the NRC to stop being a spectator and start being a regulator. The situation at River Bend and Perry can be compared to health insurance. When an insurance company finds out that one ofits customers had a pre-existing health problem that was not accurately reported on the medical survey, it will either raise the premium or cancel the policy. It will take this action because ; the risk factor is higher with the pre-existing condition. The difference between this example and the ' situation at River Bend and Perry is that the insurance company will not overlook the matter if the customer correctly spelled all of the words on the medical survey. He insurance company, unlike the NRC, focuses on substance. If the NRC carefully reviews the facts in this matter, we are confident that they will take the necessary actions to protect public health and safety. If the NRC disagrees with our contentions, we would ask that they document how each one of the many UFSAR discrepancies that we have identified in our petitions and during this presentation are satisfied. We will not accept an NRC denial of our petition based exclusively on the fact that the plants comply with their Technical Specifications, ne 2.206 petition process does not currently have an effective appeal process. So our appeal, like in the Millstone case, will be to the media, to elected officials', and to the public. I have described several safety questions that must be answered before anyone can truthfully say that nuclear plants operating with damaged fuel do not pose a threat to workers and the general public. At this , point, no one knows. Ignorance may be bliss, but bliss is not safe. What is really needed is a bliss reduction program.
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e ANSWERS TO PUBLIC PARTICIPATION QUESTIONS AND COMMENTS RAISED DURING INFORMAL PUBLIC HEARING CONCERNING 10 CFR 2.206 PETITION HELD ON FEBRUARY 22,1999
~ Question: (Ms. Connie Kline) "My first question is, isn't it true that failed fuel or defective fuel cannot be dry casked, and also cannot - could not be shipped to a repository if one existed?" (See Informal Public Hearing Transcript, beginning on Page 89, Line 20)
Answer: The Spent Fuel Project Office has recently issued interim Staff Guidance (ISG-1) on the issue of damaged fuel and dry cask storage. " Damaged fuel"is defined as spent nuclear fuel with known or suspected cladding defects greater than a hairline crack or a pinhole leak. The NRC applies this definition in matters relating to spent fuel storage as well as transportation. Damaged fuel, as previously defined, can be stored in dry cask storage systems (DCSS) consistent with a licensee's 10 CFR Part 72 license. ISG-1 states that damaged fuel should be canned for storage and transportation. The purpose of canning is to confine gross fuel particles to a known, subcritical volume during off-normal and accident conditions, and to facilitate handling and retrievability. The provision for canning damaged fuel applies to both spent fuel storage and transportation. Spent fuel, with plutonium in excess of 20 curies per pacisage, in the form of debris, particles, loose pellets, and fragmented rods or assemblies must be packaged in a separate inner container (second containment system) in accordance with 10 CFR 71.63(b). The provision for double containment applies to transportation only. As proof that the fuel to be loaded is undamaged, the staff will accept, as a minimum, a review of the records to verify that the fuel is undamaged, followed by an external visual examination of the fuel assembly prior to loading for any obvious damage. For fuel assemblies where reactor records are not available, the level of proof will be evaluated on a case-by-case basis. The purpose of this demonstration is to provide reasonable assurance that the fuel is undamaged or that damaged fuel loaded in a storage or transportation cask is canned. This provision for demonstrating the condition of the fuel applies to both storage and transportation. Desian and Reaulatorv Reauirements The provisions for storage of damaged fuel will be defined during the licensing of the dry cask storage system at each facility, NUREG-1536," Standard Review Plan for Dry Cask Storage Systems" provides guidance to the staff during the review of a DCSS license under 10 CFR Part 72. The review will consider the following aspects with regard to damaged fuel:
- 1. The DCSS must maintain confinement of radioactive material within the limits of 10 CFR Part 72 and Part 20, under normal, off-normal, and credible accident conditions. [10 CFR 72.236(l)]
- 2. The spent fuel cladding must be protected during storage against degradation that leads to gross ruptures, or the fuel must be otherwise confined such that degradation of the fuel during storage will not pose operational safety problems with respect to its removal from storage. [10 CFR 72.122(h)(1)]
ENCLOSURE 6
r s 2-
- 3. Spent fuel transfer and storage systems must be designed to remain subcritical under all credible conditions. [10 CFR 72.124(a) and 72.236(c)) '
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- 4. The applicant stiould define the range and types of spent fuel or other radioactive -
materials that the DCSS is designed to store. In addition, these specifications i should include, but are not to be limited to, the type of spent fuel (i.e., boiling-water reactor (BWR), pressurized-water reactor (PWR), or both), weights of the stored materials, dimensions & configurations of the fuel, maximum allowable enrichment of the fuel before any irradiation, burnup (i.e., megawatt-days /mtu), minimum acceptable cooling time of the spent fuel before storage in the DCSS (aged at least 1 year), maximum heat designed to be dissipated, maximum number of spent fuel elements, condition of the spent fuel (i.e., intact assembly or consolidated fuel . l rods), inerting atmosphere requirements, and the maximum amount of fuel j permitted for storage in the DCSS. For DCSSs that will be used to store radioactive materials other than spent fuel, that is, activated components associated with a spent fuel assembly (e.g., control rods, BWR fuel channels), the applicant should specify the types and amounts of radionuclides, heat generation and the relevant source strengths and radiation energy spectra permitted for storage in the DCSS.
- 5. - Examine any limitations regarding the condition of the spent fuel. If damage that could be classified as a " Gross Cladding Defect" is allowed, the effects of such damage should be assessed in later sections of the SAR. If damaged rods have been removed from a fuel assembly, determine whether a need -
exists to replace them with dummy rods before loading into the cask. Note, tne presence of an additional moderator will need to be addressed in the criticality analysis in SAR Section 6.
- 6. The regulations require that the cask be designed to withstand the effects of accident conditions and natural phenomena events without impairing its capability to perform safety functions.
Question: (Mr. Jack Feterau) "The first question, can the leaking radioactive material in the core degrade or cause increased electroly s of any of the plumbing and seals in the loop? Can the released radioactivity In the ..,p degrade the alloy or material or change the atomic structure of the material in the loop, and if so, how much? And can this cause down-the-road problems?" (See Transcript, beginning on Page 94, Line 16) Answer: Materials selected during the design of the reactor, reactor coolant piping, recirculating pump seals and other related components all take into consideration the operational environment of the systems during normal and off-normal operations including the effects of fuelleaks within the limits of the plant's technical specifications. The radioactive material present in the reactor coolant as a result of the leaking fuel assemblies will have virtually no impact upon the safe operation of the plant.
7c , Question: - (David Ellison) [M]y concern in this case is that the contamination that results from the radiation released from leaking fuel rods would present a problem for us later on when
- FirstEnergy or whatever company owns the powerplant, by the time it gets to the end of its license, walks away from the plant and leaves it in the lap of the public around here,
- and what greater exposure and what greater expense we'll have to incur at that time as a result of the inattention to leaking fuel rods at this time. I would also like to point out that the dose limits that people are allowed to receive vary widely from agency to agency, and that to have charts that indicate that you're well within limits of dose exposure while not pointing out that the dose exposure for a person working in a power plant might be totally different than a person working to' decommission and get rid of that power plant are two different things, and that it's not a meaningful thing and it's misleading, and I don't really appreciate it. (See Transcript, beginning on Page 98, Line 3)-
Answer: Mr Ellison raises a number of issues concerning the decommissioning of a nuclear power plant. In order to address these issues, the following " Frequently Asked Questions" are being reprinted from the NRC web page on decommissioning: Q: Does the licensee have an NRC license even during the decommissioning process?
- A: Yes. The NRC license is not terminated until the licensee can demonstrate that it meets the criteria for site release in the regulations. This is demonstrated by a final radiation survey that is reviewed and verified by the NRC staff. In addition, the licensee must demonstrate that the facility has been dismantled in accordance with the approved license termination plan.
Q: Could the NRC require a plant to cease operations and begin decommissioning? A: The regulations allow the NRC to revoke, suspend, or modify a license in whole or in part for failure to operate a facility in accordance with the terms of the license or for violation of, or failure to observe, any of the terms and provisions of the Atomic Energy Act, regulations, license, or order of the Commission. The NRC may issue an order to a licensee to permanently cease operations. If such an order were issued, the licensee would have 60 years from the date it permanently ceased operation until the completion of decommissioning. Q: What would happen if a licensee refuses to decommission a plant that has ceased operations? A: The Commission may levy a civil penalty against the licensee. The regulations allow the NRC to obtain a court order for the payment of a civil penalty for violations of any rule, regulation, or order, or for violation of any term, condition, or limitation of any license, in addition, the Atomic Energy Act provides for the Federal Govemment to assume responsibility for decommissioning if public health and safety are jeopardized because of inactivity on the part of the licensee. Q: - What would happen if the licensee's license expires before the decommissioning process is concluded? A: ~ The decommissioning regulations state that the license for a facility that has permanently ceased operations will continue in effect beyond the expiration date to authorize possession of
. the facility until the Commission notifies the licensee in writing that the license is terminated.
During such a period of continued effectiveness, the licensee shall take actions necessary to decommission and decontaminate the facility and shall continue to maintain the fac' .o a safe condition.
4 Q: Is the choice of decommissioning alternatives a decision that is left entirely to the licensee or does the NRC help make this decision? A: The choice of the decommissioning method is left entirely to the licensee. However, the NRC would require the licensee to reevaluate its decision if the choice (1) could not be completed as described, (2) could not be completed within 60 years of the permanent cessation of plant operations, (3) included activities that would endanger the health and safety of the public by being outside of the NRC's health and safety regulations, or (4) would result in a significant impact to the environment. O: How much occupationa' dose is received by workers during decommissioning? A: The amount of occupational dose received during the decommissioning process will depend on the design and size of the facility as well as on the plans for decommissioning. A greater amount of occupational dose is anticipated to be incurred for an immediate decontamination and dismantlement than for a storage period followed by dismantlement. Estimates were given in a generic study of decommissioning (published in 1988) that ranged from 333 person-rem for a 30- year storage period to 1874 person-rem for an immediate decontamination and dismantlement. This can be compared with the 1996 annual average for an operating plant: 126 person-rem for pressurized-water reactors and 235 person-rem for boding-water reactors. The puson-rem numbers are the doses that are received by all the workers. The dose to any one worker is expected to be below the 5-rem-a-year regulatory limit, and is usually well below this limit. l l Since that study was performed, estimates for occupational dose from decommissioning range l from 591 person-rem for the Trojan nuclear plar to 1215 person-rem for Maine Yankee (includes the dose from transportation of the low-level waste (LLW) and 996 person-rem for Haddam Neck (includes the occupational dose from the transportation of the LLW). All three of these plants are using an immediate decontamination and dismantlement type of decommissioning. O: Are there limits on the amount of occupational dose that may be received? A: Yes. The regulations state that the licensee shall control tha occupational dose to individual adults to an annual limit of 5 rem (total effective dose equivalent to the entire body) or to an organ dose equal to 50 rem. There are also dose limits to the eyes, the skin, and the extremities. O: Is the safety of the public considered in the planning for and review of decommissioring? A: Yes. The safety of the public is a major concern even though the potential for hazards to the l public from the decommissioning process and potential accidents is much less than it is when i the facility is operating. ) i Q: How much dose will the public receive during the decommissioning process? A: The only exposure anticipated to the public is from the shipment of low-level waste from the site l to the low-level waste disposal site. Estimates made in a generic study of nuclear power reactor decommissioning range from 3 person-rem for a 30-year storage period to 21 person-rem for immediate decontamination and decommissioning. The estimated public dose from the Trojan ! nuclear plant decommissioning is _4.8 person-rem. The estimated dose to the public from decommissioning the Haddam Neck plant is 11 person-rem. The radiation dose is received by people who travel along the same route as the trucks that are transporting low-level waste. However, because of the variability in the timing of each shipment, the short period of time that I
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any person would be near any of the trucks, and the small dose that is allowed 6 feet from the side of a truck (10 millirem per hour), the dose that is received by any one person traveling down
~ the hig'nway or stopped for an hour at a rest stop is a very small fracticn of the annual dose that
. the person would receive from background radiation.
Q:- Who estimates what the doses are and how are these estimates made? - TA: The licensee estimates'the doses. The doses are estimated using assumptions about the - amount of. radioactive material that will be released, or the proximity of the public to the source of radiation. The doses are calculated using NRC-approved assumptions, models, and codes. The NRC may review the licensee's estimates of the doses and often recalculates the doses
- using its own assumptions for activities with the potential for significant worker exposure.
Q: What types of effluent releases are expected, and where will they enter the env?ronment? A: Three important radiation exposure pathways need to be considered in the evaluvion of the i radiation safety of normal reactor decommissioning operations: inhalation, ingestion, and - external exposure to radioactive materials. During decommissioning, inhalation is cons:dered to ~ be the dominant pathway of public radiation exposure, since exposure to radioactive sunaces and ingestion can be minimized or eliminated as radiation pathways to the public. During te transport of radioactive wastes, inhalation arid ingestion can be minimized or eliminated as radiation pathways to the public by containing the waste in a form or a container that does not allow for release to the air or water. Therefore, for transportation of radioactive waste, extemal exposure to radioactive materials is considered to be the only pathway r.. concem. Q: How much does it cost to decommission a nuclear power plant? A: The total cost of decommissioning is dependent on the sequence and timing of the various stages of the program. The minimum amounts that are required for reasonable assurance of
; funds for decommissioning are $105 million for pressurized-water reactors and $135 million for
- boiling-water reactors. These costs are in 1986 dollars and are adjusted annually, as further specified in the regulations. These are minimum amounts to show reasonable assurance, rather than estimate, of what it would cost to decommission a specific nuclear reactor.
Actual site-specific costs incurred and estimated costs of decommissioning give a better - Indication of what the process costs. The Fort St. Vrain nuclear plant, which was a 330-megawatt-electric high-temperature gas-cooled reactor, ceased power operations in 1989 i and underwent immediate decontamination and dismantlement. The decommissioning effort
, was completed in late 1996, and the license was terminated. The total cost of decommissioning was $189 million/
The cost for decommissioning the Trojan' nuclear plant (an 1130-megawatt-electric
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pressurized-water reactor) is estimated to be on the order of $210 million in 1993 dollars, which j does not include $42 million for non-radioactive site remediation or $110 million for the . ) independent spent fuel storage installation (ISFSI) and related fuel management. The Trojan j nuclear plant is also planning an immediate decontamination and decommissioning from shutdown in 1993 to license termination in 2002. The estimated cost for decommissioning the Haddam Neck nuclear plant, a 619-megawatt-
. electric pressurized-water reactor is, $344.4 million in 1996 dollars, not including $82.3 million in spent fuel storage costs (for a total of $426.7 million). The estimated cost for decommissioning Maine Yankee, an 830-megawatt-electric pressurized water reactor, is $274.9 million in 1997 dollars. This does not include costs for spent fuel management ($53.4 million) or for site !
restoration ($49.2 million), for a total of $377.6 million. ' I I m ;
{ , r- ! 0: Who makes the estimates of the decommissioning costs? - A: The licensee makes the estimates of the decommissioning costs, or hires a contractor who has : extensive experience in making these estimates. The estimates are reviewed by the NRC. O: When are the estimates of the decommissioning costs made? l A: The NRC has regulations regarding the methods used to reasonably assure that funds will be l available to decommission the facility. The NRC has specified a table of minimum aniounts required to demonstrate reasosable assurance of funds for decommissioning by reactor type and power level ($105 million for pressurized-water reactors and $135 million for boiling-water reactors in 1986 dollars). Licensees may also perform site-specific estimates that could result in cost estimates that are higher than the generic formula amounts specified in 10 CFR 50.75 (c). An estimate is made at or about 5 years preceding the projected end of operations. At this ame, power reactor licensees shall submit a preliminary decommissioning cost estimate, which includes an up to-date essessment of the major factors that could affect the cost to decommission. If the amount of money available is inadequate, the licensee has approx lmately 5 years to adjust the money in the decommissioning trust fund to ensure that appropriate funds I are available for decommissioning. An estimate is submitted at the time that the post-shuidown decommissioning activities re;* ort (PSDAR) is submitted (no later than 2 years following permanent cessation of operations). This estimate may be (1) a site-specific cost estimate that is based on the activities and schedule that i are also discussed in the PSDAR, (2) an estimate based on actual costs at similar facilities that i have undergone similar decommissioning activities, or (3) a generic cost estimate. The NRC recommends that licensees planning an immediate decontamination and dismantlement submit a site-specific cost estimate in the PSDAR; however, a more generic one would be acceptable for facilities that are submitting their PSDAR in advance of the 2-year requirement. If a storage period is planned during decommissioning, the licensee should provide a method of adjusting the cost estimate and funding throughout the duration of the storage. The regulations also require a site-specific cost estimate, within 2 years follow. permanent ; cessation of operations,if one has not already been submitted. 1 l Finally, at the time that the license termination plan is submitted (at least 2 years before the date when the license terminates), an upda';d site-specific estimate of any remaining decommissioning costs is required.
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Q: If the first estimate of decommissioning costs is made at the time that the facility is licensed, are there metMds for adjusting for inflation? A: NRC regulations provide an adjustment factor for cost escalation that takes into account escalation factors for labor, energy, and waste burial. The labor and energy escalation factors i are obtained from regional data issued by the U.S. Department of Labor's Bureau of Labor Statistics. The waste burial cost escalation factor is taken from an NRC report, " Report on Waste i Burial Charges." - 2-
p y l Q: ~ How'does the NRC ensure that the licensee will have the money when it is needed for decommissioning?
' A:' Financial assurance is provided by the following methods:
Prepayment. In this case, at the start of operations, the licensee deposits into an account enough funds to pay the decommissioning costs. The account is segregated from the licensee's
, other assets and remains outside the licensee's administrative control of cash or liquid cssets.
Prepayment may be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of govemment securities. External sinking fund. An extemal sinking fund is a fund established and maintained by setting funds aside periodically into an account segregated from licensee assets and outside the licensee's administrative control. The total amount of these funds would be sufficient to pay decommissioning costs at the time that it is anticipated that the licensee will cease operations, i An external sinking fund may be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of govemment securities. Surety method, insurance, or other guarantee method. A surety method may be in the form of a , surety bond, letter of credit, or line of credit. Any surety method or insurance used to provide 1 financial assurance must be open-ended, or if written for a specific term, such as 5 years, must be renewed automatically unless,90 days or more preceding the renewal date, the issuer notifies the Commission, the beneficiary, and the licensee of its intent to not renew. The surety or insurance must also provide that the full tice amount be paid to the beneficiary automatically
. preceding the expiration date without proof cf forfeiture if the licensee fails to providc a replacement acceptable to the Commission within 30 days after receipt of notification of cancellation, in addition, the surety or insurance must be payable to a trust established for i decommissioning costs, and the trustee and trust must be acceptable to the Commission. The )
surety method or insurance must remain in effect until the Commission has terminated the license. " . Q: . ls there any way to ensure that the licensee does not ju'st spend all of the money in the first few j years of decommissioning and have nothing left to complete the job? -
- A: The NRC has placed regulations regarding tne amount of money that can be used from the decommissioning fund at various stages of the decommissioning process. The licensee is allowed to use 3 percent of the generic amount of funds that are specified in the regulations for :
power plants based on their size and type for decommissioning planning that may occur, even while the facility is still operating. Appropriate activities include engineering design, work package preparation, and licensing activities. After submitting the certification of permanent cessation of operations and the certification that the fuel has been removed from the reactor vessel, the licensee may use an additional 20 percent of the funds for any legitimate decommissioning activities. The licensee is prohibited from using the remaining 77 percent of the generic decommissioning funds until a site-specific cost estimate is submitted to the NRC. This cost estimate must be submitted within 2 years following permanent cessation of operations.
. Q: .What would happen if the cost of decommissioning exceeds the cmount of money in the trust fund?
A: - The various cost estimates (at the time of licensing,5 ycars before anticipated shutdown, with the PSDAR submittal,2 years following shutdown, and 2 years preceding the anticipated
-termination of the license) are a method of reevaluating the decommissioning costs at various
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- ' . times and stages in the facility's life to ensure that there will be adequate funds available to
- complete the decommissioning process. If there is insufficient money in the trust fund, typically a licensee will secure a line of credit and borrow the funds to complete the decommissioning of the facility.
. Q: What would happen if the plant has an accident and there is not enough money in the
. decommissioning trust fund to complete decommissioning and cleanup after the accident? '
< A: , Licensees are required to carry insurance, which !sseparate from the decommissioning funding requirements, in an amount that would allow cleanup of the site to such a level that decommissioning could be completed with the full amount of the decommissioning trust func' Currently, $1.06 billion per operating unit is required for such insurance coverage.
O: ' Who pays for decommissioning? . A: The particular licensee that holds the license for the facility pays for decommissioning. Subject to the public utilities commission that regulated the utility, the money for decommissioning is collected as part of the price of electricity; thus the funds for decommissioning are ultimately paid by the ratepayer in the electric bill.- Q: - What contingency plans are in place to assure that decommissioning and long-term radioactive neaterial storage will be properly performed in the event of financial default of the licensee? Who
. finances decommissioning if the licensee becomes bankrupt or insolvent?
A: . The Atomic Energy Act contains provision for the Federal Govemment to assume responsibility
- for decommissioning if pubhc health and safety are jeopardized because of inability on the part a-
- of the licensee.
. Bankruptcy does not necessarily mean that a power reactor licensee will liquidate. To date, the NRC's experience with bankrupt powet reactor licensees has been that they file under Chapter 11 of the Bankruptcy Code for reorganization, not liquidation (e.g., Public Service Company of New Hampshire, El Paso Electric Company, and Cajun Electric Cooperative). In these cases, bankrupt licensees have continued to provide adequate funds for safe operation
' and decommissioning, even as bondholders and stockholders suffered losses that were often
- severe. Because electric utilities typically provide an essential service in an exclusive franchise
' area, the NRC staff believes that, even in the unlikely case of a power reactor licensee -
liquidating, its service territory and obligations, including those for decommissioning, would
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revert to another entity without direct NRC intervention. Q: What will happen if deregulation becomes a reality? How will dereguiation affect anticipated revenue and the ability to decommission? A: The NRC has issued a final policy statement on its expectations and intended approach to - nuclear power plant licen' sees as the electric utility industry moves from an environment of rate regulation toward greater competition. This policy statement was issued on August 19,1997, and published in the Federal Register. The policy statement addresses NRC concems about the
. adequacy of decommissioning funds. The statement indicates that the NRC believes that its current regulatory framework is generally sufficient to address the expected changes, but in order to remove any ambiguities in its regulatlons and address situations that may not be
. adequately covered, the Commission is considering revising 'its financial and decommissioning
' funding assurance requirements.
V .'
y Deregulation may force some licensees to separate their systems into functional areas, with their NRC-licensed nuclear plants potentially no longer being rats regulated . This would cause some licensees to cease being an " electric utility," as defined in NRC regulations. If this occurs, the NRC will require the licensees to meet the more stringent decommissioning funding assurance requirements that apply to non-electric utilities. Electric utilities are permitted to accumulate funds for decommissioning over the remaining terms of their operating licenses. NRC regulations require most other licensees to provide funding assurance for the full estimated cost of decommissioning, either through full up-front funding or by some allowable guarantee or surety mechanism. In addition, the policy statement emphasizes that the NRC retains the right to assess the timing of " commissioning trust fund deposits and withdrawals and the liquidity of decommissioning funds for licensees that no longer are subject to rate regulatory oversight. O: What publications contain the regulations for decommissioning? A: Regulations regarding decommissioning of NRC-licensed plants appear in the Code of Federal Regulations. The Code of Federal Regulations is a codification of the general and permanent rules published in the Federal Register by the executive departments and agencies of the Federal Govemment. The Code is divided into 50 titles, which represent broad areas subject to Federal regulatiun. Each title is divided into chapters; these usually bear the name of the issuing agency. Each chapter is further subdivided into parts covering specific regulatory areas. The regulations related to decommissioning of power reactors are included in Title 10 (" Energy"), Chapter I--Nuclear Regulatory Commission, e.g., Part 20, " Standards for Protection Against Radiation"; Part 50- " Domestic Licensing of Production and Utilization Facilities"; and Part 51- " Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions." The subparts related to decommissioning are 20.1402, " Radiological criteria for unrestricted use"; 20.1403, " Criteria for license termination under restricted conditions"; 20.1404, j "Altemate criteria for license termination"; 20.1405, "Public notification and public participation"; " 20.1406, " Minimization of contamination"; 50.75, " Reporting and recordkeeping for decommissioning planning"; 50.82, " Termination of license"; 51.53, " Post-construction environmental reports *; and 51.95, " Post-construction environmental impact statements." These regulations state the technical and financial criteria for decommissioning licensed nuclear facilities. They address decommissioning, planning needs, timing, funding methods, and environmental review requirements. Copies of the Code of Federal Regulations are found in the local public documerit reading rooms and often at local libraries in the reference section. They are 8vailable for purchase from the Government Printing Office by credit card at 202-512-1800, Monday through Friday,8 a.m. to 4 p.m. EST (fax 202-512-2233,24 hours a day) or by check by writing to the Superintendent of Documents, Attn: New Orders, P.O. Box 371954, Pittsburgh, PA 15250-7954. For GPO Customer Service, call 202-512-1803. For more information on the subject of decommissioning, members of the public with access to the world wide web cre encouraged to visit the NRC Web Page on " Staff Responses to Frequently Asked Questions on Decommissioning Nuclear Power Reactors" at: http1/www. arc. gov /N R C/N U R EG S/S R1628/pa rt06.html - _}}