ML111570242
| ML111570242 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 04/18/2011 |
| From: | Boska J Plant Licensing Branch 1 |
| To: | |
| Boska J, NRR/DORL, 301-415-2901 | |
| References | |
| G20110218, OEDO-2011-0223, TAC ME5930, TAC ME5931 | |
| Download: ML111570242 (49) | |
Text
Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION
Title:
10 CFR 2.206 Petition Review Board RE Indian Point Nuclear Generating Unit Docket Number:
50-247 and 50-286 Location:
(telephone conference)
Date:
Monday, April 18, 2011 Work Order No.:
NRC-841 Pages 1-48 Edited by John Boska, NRC Petition Manager NEAL R. GROSS AND CO., INC.
Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.
Washington, D.C. 20005 (202) 234-4433
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 1
UNITED STATES OF AMERICA 1
NUCLEAR REGULATORY COMMISSION 2
+ + + + +
3 10 CFR 2.206 PETITION REVIEW BOARD (PRB) 4 CONFERENCE CALL 5
RE 6
INDIAN POINT FUEL PEAK CLAD TEMPERATURE 7
DOCKET NOS. 50-247 + 50-286 8
+ + + + +
9 MONDAY 10 APRIL 18, 2011 11
+ + + + +
12 The conference call was held, Fred Brown, 13 Chairperson of the Petition Review Board, presiding.
14 PETITIONERS:
15 PHILIP MUSEGAAS AND MARK LEYSE, representing 16 RIVERKEEPER 17 PETITION REVIEW BOARD MEMBERS:
18 FRED
- BROWN, Chair,
- Director, Division of 19 Inspection and Regional Support, NRR 20 JOHN BOSKA, Project Manager, NRR 21 TANYA MENSAH, PRB Coordinator, NRR 22 BRICE BICKETT, Region I Division of Reactor 23 Projects 24 RICHARD DUDLEY, Rulemaking Branch, NRR 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 2
SHIH-LIANG WU, Nuclear Performance and Code 1
Review Branch, NRR 2
3 PRB ADVISORS:
4 CHRISTOPHER HOTT, Office of Enforcement 5
BRETT KLUKAN, Office of General Counsel 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 P-R-O-C-E-E-D-I-N-G-S 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 3
1:01 p.m.
1 MR. BOSKA: Good afternoon. I'd like to 2
thank everyone for attending this meeting. My name is 3
John Boska, and I am the NRC petition manager for this 4
petition.
5 We're here today to allow the petitioners, 6
Ms. Brancato and Mr. Leyse, to address the Petition 7
Review Board on behalf of Riverkeeper concerning their 8
2.206 petition dated March 28th, 2011, on the fuel 9
peak cladding temperature at Indian Point Nuclear 10 Generating Unit Numbers 2 and 3, which are located 11 about 24 miles north of New York City on the east 12 bank of the Hudson River.
13 I
am the petition manager for the 14 petition. The Petition Review Board chairman is Fred 15 Brown.
16 As part of the Petition Review Board's 17 review of this petition, Ms. Brancato and Mr. Leyse 18 have requested this opportunity to address the 19 Petition Review Board, which may also be referred to 20 as the PRB.
21 This meeting is scheduled to conclude by 22 3:00 pm. The meeting is being recorded by the NRC 23 Operations Center, and will be transcribed by a court 24 reporter. The transcript will become a supplement to 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 4
the petition. The transcript will also be made 1
publicly available.
2 I'd like to open this meeting with 3
introductions. As we go around the room, please be 4
sure to clearly state your name, your position, and 5
the office that you work for within the NRC for the 6
record.
7 I'm John Boska. I'm a project manager in 8
the office of Nuclear Reactor Regulation, which is 9
also referred to as NRR.
10 MS. SALGADO: This is Nancy Salgado. I'm 11 a branch chief in the division of Operating Reactor 12 Licensing, NRR.
13 MR.
WU:
Shih-Liang Wu, Nuclear 14 Performance and Code Review Branch, NRR.
15 MR. KLUKAN: Brett Klukan. I'm the Office 16 of General Counsel advisor to the PRB.
17 MR. BROWN: Fred Brown, Director of 18 Division of Inspection and Regional Support within the 19 Office of NRR.
20 MR. DUDLEY: Richard Dudley. I'm a 21 rulemaking project manager from the NRR rulemaking 22 branch.
23 MS.
MENSAH:
Tanya
- Mensah, 2.206 24 coordinator from the Division of Policy and Rule 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 5
Making, NRR.
1 MR. HOTT: Chris Hott. I'm an enforcement 2
specialist in the Office of Enforcement.
3 MR.
BOSKA:
We have completed 4
introductions in the room.
5 At this time, are there any other NRC 6
participants from NRC headquarters on the phone?
7 Hearing
- none, are there any NRC 8
participants from the Regional Office on the phone?
9 MR. BICKETT: Yes, this is Brice Bickett 10 from NRC Region 1. I'm a Senior Project Engineer.
11 MR. BOSKA: Thank you, Brice.
12 Are there any representatives for the 13 licensee on the phone?
14 MR. WALPOLE: Bob Walpole, Indian Point.
15 MR. IRANI: Adi Irani, headquarters.
16 MR. BOSKA: All right. Thank you.
17 Ms. Brancato and Mr. Musegaas, would you 18 please introduce yourself for the record?
19 MR. MUSEGAAS: Sure. This is Philip 20 Musegaas. I'm an attorney and the Hudson River 21 Program Director at Riverkeeper.
22 And John, I apologize for any confusion.
23 I will be giving our statement today on behalf of 24 Riverkeeper. And I'm joined by Mark Leyse, but I'll 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 6
be on this call instead of Deborah Brancato. So 1
again, sorry for any confusion there.
2 MR. BOSKA: All right. That's fine.
3 MR. MUSEGAAS: Thank you.
4 MR. BOSKA: Thank you, Mr. Musegaas.
5 Mr. Leyse, would you please introduce 6
yourself for the record?
7 MR. M. LEYSE: Sure. Mark Leyse.
8 MR. BOSKA: All right. It is not required 9
for members of the public to introduce themselves for 10 this call. However, if there are any members of the 11 public on the phone that wish to do so, please state 12 your name for the record.
13 MR. LOCHBAUM: This is David Lochbaum, 14 Director of the Nuclear Safety Project for the Union 15 of Concerned Scientists.
16 MR. SIM: Hi, I am Bob Sim (phonetic),
17 Assistant Attorney General, New York.
18 MR. BOSKA: All right. Welcome.
19 MR. R. LEYSE: Robert Leyse, citizen.
20 MR. WALD: My name is Matt Wald. I'm a 21 reporter at the New York Times. I'm not a 22 participant, I'm just listening. Thank you.
23 MR. BOSKA: All right. Thank you for the 24 introductions.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 7
I'd like to emphasize that we each need to 1
speak clearly and loudly to make sure that the court 2
reporter can actually transcribe this meeting. If you 3
do have something that you'd like to say, please first 4
state your name for the record.
5 For those dialing into the meeting, please 6
remember to mute your phones to minimize any 7
background noise or distractions. If you do not have 8
a mute button, this can be done by pressing the keys 9
star six. To un-mute your phone, press the star six 10 keys again.
11 Please do not place this call on hold, 12 since many phone systems play music when a call is on 13 hold, which is very annoying for the other callers.
14 Thank you.
15 Next, I'd like to share some background on 16 our process. Section 2.206 of Title 10 of the Code of 17 Federal Regulations describes the petition process, 18 the primary mechanism for the public to request 19 enforcement action by the NRC in a public process.
20 This process permits anyone to petition 21 the NRC to take enforcement-type action related to NRC 22 licensees or licensed activities. Depending on the 23 results of its evaluation, the NRC could modify, 24 suspend, or revoke an NRC-issued license or take any 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 8
other appropriate enforcement action to resolve a 1
problem.
2 The NRC staff's guidance for the 3
disposition of 2.206 petition requests is in 4
Management Directive 8.11, which is publicly 5
available.
6 The purpose of today's meeting is to give 7
the petitioners an opportunity to provide any 8
additional explanation or support for the petition 9
before the Petition Review Board makes an initial 10 recommendation on whether or not to accept this 11 petition for review.
12 The Petition Review Board typically 13 consists of a chairman, usually a manager at the 14 senior executive service level at the NRC. It has a 15 petition manager and a PRB coordinator.
16 Other members of the board are determined 17 by the NRC staff, based on the content of the 18 information in the petition request.
19 At this time, I would like to introduce 20 the Board. Fred Brown is the Petition Review Board 21 chairman. I am the petition manager for the petition 22 under discussion today. Tanya Mensah is the office's 23 PRB coordinator.
24 Our technical staff includes Shih Liang-Wu 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 9
from the NRR's Nuclear Performance and Code Review 1
Branch, Brice Bickett from the NRC's Region 1 Division 2
of Reactor Projects, and Richard Dudley, from NRR's 3
Rulemaking Branch.
4 We also have legal advice from Brett 5
Klukan from the NRC's Office of General Counsel, and 6
advice from Christopher Hott from the Office of 7
Enforcement.
8 As described in our process, the NRC staff 9
may ask clarifying questions in order to better 10 understand the petitioner's presentation and to reach 11 a reasoned decision whether to accept or reject the 12 petitioner's request for review under the 2.206 13 process.
14 I would like to summarize the scope of the 15 petition under consideration, and the NRC activities 16 to date.
17 On March
- 28th, 2011, Ms.
Brancato 18 submitted to the NRC a petition under 10 CFR 2.206 19 regarding the fuel peak cladding temperature at Indian 20 Point Nuclear Generating Unit numbers 2 and 3, which 21 may also be called IP-2 and IP-3.
22 This petition is available from the NRC's 23 public website, www.nrc.gov, from the electronic 24 reading room under the ADAMS documents with the 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 10 accession number ML110890956.
1 The petitioners request that the NRC order 2
the licensee of IP-2 and IP-3 to lower the licensing 3
basis peak cladding temperatures of IP-2 and IP-3 in 4
order to provide necessary margins of safety to help 5
prevent partial or complete meltdowns in the event of 6
loss-of-coolant accidents, also called LOCAs.
7 The petitioners state that experimental 8
data demonstrates that IP-2 and IP-3's licensing basis 9
peak cladding temperatures of 1,937 degrees Fahrenheit 10 and 1,961 degrees Fahrenheit, respectively, do not 11 provide necessary margins of safety in the event of 12 LOCAs.
Such data demonstrates that IP-2's and IP-13 3's licensing basis peak cladding temperatures need to 14 be decreased to temperatures lower than 1,832 degrees 15 Fahrenheit in order to provide necessary margins of 16 safety.
17 Second, the petitioners request that the 18 NRC order the licensee of IP-2 and IP-3 to determine 19 how far below 1,832 degrees Fahrenheit the licensing 20 basis peak cladding temperature values of IP-2 and IP-21 3 need to be lowered in order to provide necessary 22 margins of safety.
23 Third, the petitioners request that the 24 NRC order the licensee of IP-2 and IP-3 to lower both 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 11 of IP-2 and IP-3's licensing basis peak cladding 1
temperatures to 1,600 degrees Fahrenheit until 2
conservative values for IP-2 and IP-3 are determined.
3 Fourth, the petitioners request that the 4
NRC order the licensee of IP-2 and IP-3 to demonstrate 5
that IP-2 and IP-3 emergency core cooling systems, 6
also called ECCS, will effectively quench the fuel 7
cladding in the event of LOCAs and prevent partial or 8
complete meltdown. Experimental data indicates that 9
IP-2 and IP-3's ECCS may not effectively quench the 10 fuel cladding in the event of LOCAs, if fuel cladding 11 temperatures approached or reached IP-2 and IP-3's 12 licensing basis peak cladding temperatures of 1,937 13 degrees Fahrenheit and 1,961 degrees Fahrenheit, 14 respectively.
15 The petitioners also state that, although 16 revisions to the 10 CFR 50.46(b)(1) of 2200 degrees 17 Fahrenheit on peak cladding temperatures have been 18 proposed in a rule-making petition, PRM-50-93, this 19 petition has been filed separately under 10 CFR 2.206 20 because the concerns affect IP-2 and IP-3 and need 21 prompt resolution to protect the lives, property, and 22 environment of the people of New York. The safety 23 issues raised in this petition are of an immediate 24 nature and require prompt NRC review and action.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 12 Allow me to discuss the NRC activities to 1
date. On March 28th, 2011, the NRC received this 2
petition. On April 6th, the petition manager 3
contacted the petitioners to offer the opportunity to 4
address the PRB, to which the petitioners agreed.
5 That led to this teleconference.
6 As a reminder for the phone participants, 7
please identify yourself if you make any remarks, as 8
this will help us in the preparation of the meeting 9
transcript that will be made publicly available.
10 Thank you.
11 At this time, I'll turn it over to the 12 Petition Review Board chairman, Fred Brown.
13 MR. BROWN: Good afternoon. Welcome to 14 this meeting regarding the 2.206 petition submitted by 15 Riverkeeper.
16 First, I would like to explain the purpose 17 of this meeting. This meeting is not a hearing, nor 18 is it an opportunity for the petitioners to question 19 or examine the PRB on the merits or the issues 20 presented in the petition request.
21 No decisions regarding the merits of this 22 petition will be made at this meeting. Following this 23 meeting, the Petition Review Board will conduct its 24 internal deliberations. The outcome of this internal 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 13 meeting will be discussed with the petitioners.
1 At this time, I'd like to turn over to 2
Riverkeeper the presentation to allow you to provide 3
any information you believe the PRB should consider as 4
part of this petition, especially reasons why this 5
petition should be considered separately from the 6
petition for rulemaking.
7 You may proceed.
8 MR. MUSEGAAS: Thank you, Mr. Brown.
9
- Again, this is Phillip
- Musegaas, representing 10 Riverkeeper here at this PRB. I will make a brief 11 opening statement, probably no more than five or ten 12 minutes, and then I'll turn it over to Mark Leyse, who 13 will explain and give a presentation about the 14 technical basis for the petition.
15 I want to start out by thanking the NRC 16 for setting up this meeting, and giving us an 17 opportunity to present information on the petition.
18 Just as a little bit of a review, 19 Riverkeeper is a non-profit, membership-supported 20 organization.
We're a
501(c)(3) environmental 21 organization. We've been working on the Hudson River, 22 protecting the Hudson River from pollution for about 23 the past 40 years in one form or another.
24 We have been involved with Indian Point 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 14 issues, initially with environmental concerns when the 1
plant was first licensed beginning in 1972, and then 2
subsequently, after the terrorist attacks of September 3
11th, Riverkeeper filed a 2.206 petition related to 4
safety and security issues at that time, and have been 5
involved since 2001 on a range of safety and Nuclear 6
Regulatory Commission issues related to Indian Point 7
since that time.
8 Riverkeeper also petitioned to intervene 9
in the NRC's license renewal proceeding for Indian 10 Point in 2007.
11 Riverkeeper has significant current safety 12 concerns about the operation of Indian Point at its 13 current operating level.
14 And thank you, Mr. Boska, for spelling out 15 our request for relief. I was going to do that, but I 16 appreciate you reiterating that, and so I won't repeat 17 that since you did that already.
18 And so I will discuss, as you requested, 19 why Riverkeeper believes that this petition under 20 section 2.206 should be accepted and why it's 21 distinguishable from the rulemaking petition that has 22 already been submitted.
23 The main concern here is that because 24 Indian Point is operating at the level described in 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 15 the
- petition, fuel temperatures at the levels 1
described in the petition, we have concerns that if 2
there was a loss of coolant accident and oxidation 3
occurred and a
meltdown
- occurred, that the 4
consequences of an accident would be severe.
5 And so I'm going to focus on the location 6
of Indian Point, what we consider the enhanced risks 7
of the plant's operation, given recent information, 8
and, of course, the consequences of an accident.
9 Indian Point, as you mentioned already, is 10 located approximately 25 miles from New York City, 11 from the Bronx borough of New York City. Indian 12 Point's about 34 miles from Times Square, which is 13 roughly midtown Manhattan.
14 And within 10 miles of Indian Point, there 15 are 300,000 people. Within 50 miles of the plant, 16 there are approximately 17 to 20 million people, 17 either living, or living and working, in that area.
18 Recently we have learned, over the past 19 couple of years, new information about enhanced 20 seismic risk at Indian Point. There is a 2008 21 Columbia University study done by a Lamont-Doherty 22 Earth Institute seismologist, that concluded that in 23 addition to the Ramapo fault that was already known to 24 be near Indian Point, there was an additional fault 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 16 line.
1 And in addition, by looking at the history 2
of earthquakes in the area, the Columbia seismologist 3
concluded that Indian Point was, in fact, at risk of 4
an earthquake up to 7.0 on the Richter scale.
5 Entergy, the current owners of Indian 6
Point, have said publicly that they believe the two 7
reactors could withstand up to a 6.1, I believe, on 8
the Richter scale.
9 So we have significant concerns that the 10 plant may not be built or designed to withstand the 11 maximum earthquake that could be experienced in this 12 area. And that is fairly new information.
13 This information from Columbia has not 14 been considered or not been assessed to our knowledge 15 by the Nuclear Regulatory Commission. In addition to 16 the Columbia study, there is also a September 2010 17 Nuclear Regulatory Commission seismic risk study, 18 which places Indian Point 3 at an increased risk of 19 core damage under the core damage frequency 20 calculations from an earthquake. And that information 21 has not been considered in the license renewal process 22 for Indian Point or in any other formal regulatory 23 process that we are aware of. And so that is an 24 outline of the seismic risks that we are concerned 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 17 about.
1 In addition to this, I've spoken about 2
location, I've spoken about risk in terms of the 3
seismic damage or the potential for an earthquake 4
that's beyond the design basis of the plant.
5 I would also like to talk about the 6
consequences again. And in our petition, Riverkeeper 7
included and also cited, too, a 2004 report we 8
commissioned from the Union of Concerned Scientists, 9
Dr. Ed Lyman.
10 The report is called "Chernobyl on the 11 Hudson," and that report describes how a loss of 12 coolant accident could lead to extremely catastrophic 13 near-term and long-term fatalities and economic and 14 property damage in the area that would be affected by 15 a significant radiological release from Indian Point.
16 And so, taken together, we have the 17 location of Indian Point, we have the seismic risk, we 18 have the consequences of an accident spelled out in an 19 independent report.
20 Finally, I would just like to comment on 21
- the, and this relates back to
- location, 22 Riverkeeper's ongoing concerns about the feasibility 23 of evacuating both the 10-mile area around the plant, 24 which is the designated emergency planning zone, as 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 18 well as the absolute lack of feasibility of evacuating 1
a larger 50-mile zone around the nuclear plant.
2 And these comments are given in light of 3
what has happened at Fukushima, Japan, and the NRC's 4
recommendations to Americans living in Japan to 5
evacuate to 50 miles from that plant.
6 And on that
- note, I
will end my 7
preliminary comments and turn it over to Mark Leyse.
8 And I may give a few additional comments at the end of 9
our presentation. But for now, I'll turn it over to 10 Mark to give his presentation of the technical basis.
11 Thank you.
12 MR. M. LEYSE: Thank you, Phillip.
13 I first wanted to see, because I'm going 14 to probably be maybe half an hour or so, I wanted to 15 see if anyone else wanted to say anything before I 16 started.
17 Okay. In such case, I will proceed.
18 One thing, just before I begin, I want to 19 point out that the scenario we're discussing is a 20 fast-moving
- accident, where Fukushima, I
would 21 classify as a slow-moving accident.
22 What we're talking about, you could have a 23 partial meltdown that is well underway within ten 24 minutes time after a large pipe break, but I will get 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 19 back to that.
1 And first, I want to start out -- that in 2
the petition, I think the Petition Review Board should 3
look at page 34, there's a statement that says that 4
the Atomic Energy Commission, in responses to 5
questions submitted by Anthony Z. Roisman, this was in 6
the original Indian Point Unit 2 licensing hearing, 7
Atomic Energy Commission stated that the calculated 8
metal water reaction is negligible below 1900 degrees 9
Fahrenheit.
10 Now, that is one of the major premises on 11 which the original Unit 2 was licensed. And I want to 12 point out that there is data from thermal hydraulic 13 experiments that demonstrates that the zirconium-steam 14 reaction is not negligible below 1900 degrees 15 Fahrenheit.
16 I will talk about a test, it's from 17 Thermal-Hydraulic Experiment 1, it's TH-1, and that's 18 test number 130.
19 TH-1 test 130 was driven by small amounts 20 of fission heat, such that it would simulate decay 21 heat that would occur during a loss of coolant 22 accident.
23 In TH-1 test number 130, the reactor shut 24 down when the peak cladding temperature was 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 20 approximately 1,850 degrees Fahrenheit. And after the 1
reactor shut
- down, cladding temperatures kept 2
increasing because of the heat that was generated from 3
the metal-water reaction.
4 And the peak measured cladding temperature 5
was 2,040 degrees Fahrenheit, so there would have been 6
a very small amount of heat that would have been added 7
from actual decay heat.
8 I
just want to
- explain, there was 9
originally a very small amount of fission heat to 10 simulate decay heat, so they're running at very low 11 power, actually 0.37 kilowatts per foot, so there 12 would have been, say, 5 percent of that value. But 13 that wouldn't be enough to push the cladding 14 temperature up 190 degrees Fahrenheit after the 15 reactor shut down.
16 So, data from thermal-hydraulic 17 experiments demonstrates that the zirconium-steam 18 reaction is not negligible below 1900 degrees 19 Fahrenheit, and that was one of the original premises 20 on which Indian Point Unit 2 was licensed.
21 And that still holds true today, because 22 what they had back then was the Baker-Just 23 correlation. And if the Baker-Just correlation would 24 calculate that the metal-water reaction is negligible 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 21 below 1,900 degrees Fahrenheit, so would a best-1 estimate metal-water correlation. What they use at 2
Indian Point now is best estimate, and that's like 3
Cathcart-Pawel.
4 So, this is a very serious problem. You 5
have experimental data -- this was from an experiment 6
that was conducted in the early 1980s, and you did 7
nothing.
8 You have not examined that. You're still 9
allowing these plants close to New York City to 10 operate, and you've done nothing to address the 11 problem that actual experimental data demonstrates 12 that the metal-water reaction is not negligible below 13 1,900 degrees Fahrenheit.
It's actually very 14 substantial below 1,900 degrees Fahrenheit.
15 Another thing is, that information from 16 those tests, the TH-1 tests, in 2005, NRC stated that 17 it was actually reviewing that data from the TH-1 18 tests to determine its value for assessing the current 19 generation of codes, that would be computer models 20 such as TRAC-M, now renamed TRACE.
21 So in 2011 dollars, billions of dollars 22 have been spent on LOCA research, and I think it's 23 about time that NRC fixed the flaws in its LOCA 24 evaluation models.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 22 And currently, the heat transfer 1
coefficients used in LOCA evaluation models are still 2
not based on tests that were actually conducted with 3
zirconium alloy bundles.
4 What you have are tests that were 5
conducted with stainless steel bundles, and I will get 6
back to it, but it's just an observation, it seems 7
that NRC really does not like to conduct experiments 8
with zirconium alloy bundles.
9 First, you have the metal-water reaction 10 correlations. Those are based off of very tiny tube 11 specimens instead of being based off of zirconium 12 alloy bundles. And there is experimental data that 13 shows that the reaction rates are different.
14 And then, when it comes to the heat 15 transfer coefficients, it's the same thing. You don't 16 have the zirconium alloy bundle. Instead, you use a 17 stainless steel bundle, and stainless steel just does 18 not oxidize. And the oxidation just doesn't generate 19 heat like a zirconium alloy bundle does.
20 And yet, you have all this data, and 21 you've done nothing to improve your computer models, 22 and you're allowing Indian Point to operate in the 23 face of all of this data that shows that there are 24 major flaws in your models.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 23 Now, next, I'm going to move on to 1
something that I point out, I think this section 2
starts on page 24 of the petition. It's regarding 3
Entergy's best estimate ECCS evaluation calculations.
4 I want to point out that there is a report 5
from the Nuclear Energy Agency Group of Experts, OECD, 6
Nuclear Energy Agency, the title of the report is, 7
"In-Vessel and Ex-Vessel Hydrogen Sources." I'm 8
quoting from part one, "GAMA perspective statement on 9
in-vessel hydrogen sources." This was published in 10 2001, ten years ago.
11 And I quote from that, "Until recently, 12 the experimental data base on quenching phenomenon was 13 rather scarce. The available zircaloy-steam oxidation 14 correlations were not suitable to determine the 15 increased hydrogen production in the few available 16 tests, CORA, LOFT LP-FP-2."
17 Now, in the petition that Riverkeeper 18 submitted, there is a lengthy discussion of the CORA 19 tests and also the LOFT LP-FP-2 experiment. Now, 20 those are -- they are large, integral experiments, 21 well, especially LOFT LP-FP-2. That's a large 22 integral experiment.
23 And you just get a different metal-water 24 reaction rate than you do from your Baker-Just or 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 24 Cathcart-Pawel, which are based on experiments with 1
tiny tube specimens, like one inch long, very small.
2 Anyway, because there were problems with 3
these metal-water correlations for determining the 4
increased hydrogen production in experiments like LOFT 5
LP-FP-2, we argue that this indicates that the 6
available zircaloy oxidation kinetics
- models, 7
including best estimate models used at Indian Point 8
are non-conservative for use in analyses that 9
calculate the metal-water reaction rates that would 10 occur in the event of a LOCA.
11 So, this OECD Nuclear Energy Agency paper 12 was published in 2001. It explicitly states that 13 there are problems with these correlations when 14 they're applied to integral experiments.
15 And the NRC has done nothing, again, 16 absolutely nothing. In fact, in 2002, Robert H.
17 Leyse, submitted a petition. It was called PRM-50-76.
18 And that petition addressed the fact that the Baker-19 Just and Cathcart-Pawel correlations are deficient 20 because they were not developed to consider how heat 21 transfer would affect zircaloy-steam reaction kinetics 22 in the event of a LOCA.
23 And NRC denied that petition in 2005, and 24 one of the reasons they used to justify -- one of 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 25 their statements was that there was no data available 1
to substantiate Robert H. Leyse's claims.
2 Meanwhile, you know, there was this paper 3
published in 2001, but apparently that didn't seem to 4
matter, even though it was an OECD Nuclear Energy 5
Agency paper.
6 Now, I just want to mention just to kind 7
of examine in more detail what NRC said in its denial 8
of PRM-50-76. It's -- now I'm quoting. This is NRC's 9
statement.
10 "For the development of oxidation 11 correlations, limited by oxygen diffusion into the 12 metal where well-characterized isothermal -- " (that's 13 holding the temperature of the specimen constant,) " -
14
-tests are more important than the complex thermal-15 hydraulics suggested by Robert H. Leyse."
16 "Robert H.
Leyse's suggested use of 17 complex thermal-hydraulic conditions would be counter-18 productive in reaction kinetics studies, because 19 temperature control is required to develop a
20 consistent set of data for correlation development."
21 "Isothermal tests allow this needed 22 temperature control. It is more appropriate to apply 23 the developed correlations to more prototypic 24 transients, including complex thermal-hydraulic 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 26 conditions to verify that the proposed phenomena 1
embodied in the correlations are indeed limiting."
2 "This was what was done by Westinghouse in 3
WCAP 7665," (that's the FLECHT report), "and by 4
Cathcart-Pawel in NUREG-17, and by the NRC in its 5
technical safety analysis of PRM-50-76."
6 Now, I know I'm repeating myself, but 7
again, here you had an integral test, and it -- like I 8
said, published in 2001 in an OECD Nuclear Agency 9
Paper, and it explicitly stated that the available 10 zirconium alloy steam oxidation correlations were not 11 suitable to determine the increased hydrogen 12 production in the few available tests, including the 13 LOFT LP-FP-2 experiment.
14
- Now, I
just want to touch upon 15 Westinghouse's analysis of WCAP 7665. This would be 16 the analysis of the four zircaloy bundle runs that 17 were conducted in the FLECHT experiments. These are 18 very important experiments.
19 The Appendix K to Part 50 is still based, 20 for reflood heat transfer coefficients, that's based 21 on the test with stainless steel fuel rods from WCAP 22 7665.
23 But anyway, back to the zirconium bundle 24 runs, four of them, I point out that in the petition 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 27 that there was no -- well, I'm not sure if I mentioned 1
it in the petition, but Ive pointed it out, there was 2
no metallurgical data from the locations of run 9573 3
that incurred runaway oxidation, because Westinghouse 4
did not obtain such data.
5 So neither Westinghouse nor the NRC 6
applied the Baker-Just correlation to metallurgical 7
data from locations of run 9573 that incurred runaway 8
oxidation.
9 And in its analyses, NRC did not apply the 10 Cathcart-Pawel oxygen uptake and zirconium dioxide 11 thickness equations to metallurgical data from those 12 locations since, as I said, those locations were not 13 measured.
14 And furthermore, it's reasonable to assume 15 that, as in the CORA-2 and CORA-3 experiments in which 16 local steam starvation conditions are postulated to 17 have occurred, during FLECHT run 9573 the violent 18 oxidation essentially consumed the available steam, so 19 that the time-limited and local steam starvation 20 conditions, those that cannot be detected in a post-21 test investigation, those would have occurred.
22 Now, those would have occurred at the 23 locations where runaway oxidation did not occur. But 24 as I
pointed
- out, Westinghouse didn't take 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 28 measurements from the parts that actually incurred the 1
runaway oxidation for that one bundle. They measured 2
the parts that would have been steam-starved, and that 3
is just not a legitimate verification of the adequacy 4
of Baker-Just and Cathcart-Pawel for use in ECCS 5
evaluation calculations.
6 And I think just the fact that NRC didn't 7
really look into what was really going on, the fact 8
that they weren't analyzing sections of that bundle 9
that did incur runaway oxidation, and they were 10 actually evaluating sections that would have most 11 likely been steam starved.
12 Now, I want to point out that here we are 13 talking about problems with metal-water reaction 14 correlations for use in analyses that would predict 15 the metal-water reaction rates that would occur in the 16 event of a LOCA. But this was actually, this very 17 same issue was raised about 40 years ago by Union of 18 Concerned Scientists during the original licensing 19 hearing for IP-2.
20 Dan Ford of UCS, for example, he pointed 21 out that the Baker-Just correlation was, and I quote, 22 "derived from experimental data that is completely 23 outside of the context of nuclear systems," end of 24 quote.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 29 So that means, its derived from single-1 rod separate effects tests, and he also pointed out 2
that this metal-water reaction correlation had not 3
been derived from integral, you know, large-scale 4
integral tests.
5 And that's in the petition, in a section 6
that starts at page 21. So that gives you some 7
history, that this has been going on for quite a 8
while, criticisms of this type of problem.
9 And I want to point out that Baker-Just, 10 again, Baker-Just is not used at Indian Point now.
11 What they presently use are best estimates, and that 12 also involves the Cathcart-Pawel correlation.
13 Now, just to point out, Cathcart-Pawel, 14 that correlation is based on data conducted with 15 experiments in two different furnaces. And in one, 16 the specimen was 18 inches long, but actually, only a 17 small segment of that tube in close proximity to the 18 thermocouple station served as the specimen. The 19 other -- the full 18-inch length, part of that was to 20 just hold it in place.
21 And then, in a different furnace, the 22 specimen was about 1.2 inches long. So, you know, is 23 it any wonder that you'd get these metal-water 24 reaction correlations derived from such small tube 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 30 experiments, and then when you do a larger-scale test, 1
like LOFT LP-FP-2, these correlations, -- they're not 2
suitable for analyses.
3 And the increased hydrogen production that 4
-- in a test like LOFT LP-FP-2 -- these correlations 5
fall short of predicting that.
6 So, again, LOCA model evaluations, the 7
models are flawed, and there's data out there that 8
demonstrates that they're inadequate for use in 9
analyses.
10 The metal-water reaction correlations are 11 inadequate for use in analyses that predict the 12 zircaloy oxidation rates that would occur in a LOCA, 13 and yet, the NRC still does not do anything to fix the 14 problem.
15 And now, I want to describe the type of 16 accident that we're talking about in the petition. As 17 I had mentioned before, Fukushima is what I would call 18 a slow-moving accident, as opposed to this, which is a 19 fast-moving accident.
20 Once there was a loss of coolant accident 21 at either plant, within about 60 seconds, the peak 22 cladding temperatures, they would start approaching 23 temperatures up around 1,000 degrees Celsius, around 24 1,832 degrees Fahrenheit. That would happen within 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 31 about 60 seconds' time.
1 And what could happen at that point is 2
that in a local area, the fuel rods could incur 3
runaway oxidation, and then, within another 60 seconds 4
or so, there would be a rapid temperature escalation, 5
and it could get up to about 3300 degrees Fahrenheit, 6
where the -- that's where zircaloy starts to melt, or 7
zirconium alloy starts to melt. I believe the Indian 8
Point plants actually have Zirlo in them, but that is 9
also a zirconium alloy.
10 And so within a couple of minutes also 11 there would be extensive core damage. And this is all 12 depicted in Appendix F of the petition. There's a 13 graph there that shows the progression of a serious 14 accident and how this is very rapid.
15 And this is all based off of data like the 16 LOFT LP-FP-2 experiment and also the CORA experiments, 17 in which there was data that recorded these rapid 18 temperature escalations, commencing at temperatures as 19 low as 1,832 degrees Fahrenheit.
20 So just what happens, again, just in a 21 little more detail, the loss of coolant accident, so 22 the fuel is uncovered, and they're trying to pump 23 water back into the core. There can be problems with 24 pressurized water reactors, there can be steam 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 32 binding. So we're arguing that the models used at 1
Indian Point for Unit 2 and 3, the predicted maximum 2
temperatures of the fuel rods is 1,937 degrees 3
Fahrenheit and 1,961 degrees Fahrenheit respectively.
4 That's what's predicted.
5 But we're saying that as the fuel rods 6
approach those temperatures, all taking place within 7
under a minute, it could incur runaway oxidation. And 8
then that is going to drive the temperature in the 9
local area up to around 3,300 degrees Fahrenheit.
10 Meanwhile, as they're rising up, 11 increasing rapidly in temperature, there are also 12 other assembly components, like the control rods, 13 they're going to be rapidly moving up behind them, 14 maybe around 400 degrees Fahrenheit behind them.
15 And then when the stainless steel cladding 16 or casing of the control rods, when that gets up to 17 around 2,200 or so Fahrenheit, it can have eutectic 18 reactions with
- Zircaloy, and that can cause 19 liquefaction at those temperatures. This was observed 20 on cameras in the CORA experiments.
21 And then you can also have a problem at 22 that temperature, the inside of the control rod, which 23 is silver, indium, and cadmium, that has melted at a 24 lower temperature. But when it gets up to around 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 33 2,200 Fahrenheit, that can burst, and that can spray 1
out onto zircaloy and also dissolve it rather rapidly, 2
again, observed on cameras.
3 So we're talking about a
- great, 4
substantial amount of damage that can occur within 5
just a couple minutes' time. Again, this is in 6
Appendix F of the petition.
7 So in under ten minutes' time, there would 8
be a substantial amount of hydrogen generation from 9
this, hydrogen from the zirconium alloy oxidation, 10 hydrogen from the oxidation of stainless steel at 11 higher temperatures, but primarily from the zirconium 12 alloy. And there would also be a substantial amount 13 of hydrogen in the containment building.
14 And now, I want to just point out and 15 refer to an article -- or, not an article, it was an 16 entry, a New York Times entry that Matt Wald wrote.
17 It was on the Green, a blog about energy and the 18 environment, and dated March 31st, 2011.
19 And there is a quote from the Indian Point 20 spokesman, James F. Steets. So I just want to read 21 from one sentence from Matt Wald's article.
22 "James F. Steets, a spokesman for Indian 23 Point, said that Units 2 and 3 there each had two 24 recombiners, and that one alone could eliminate all 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 34 the hydrogen in a major accident."
1 So I just want to point out that according 2
to one paper, what Steets said is just simply 3
incorrect. Now, again, this is just one paper that I 4
have, but I want to read from that. It's Nuclear 5
Engineering and Design, Number 230, published in 2004.
6 The title is, Studies on Innovative 7
Hydrogen Recombiners as Safety Devices in the 8
Containments of Light Water Reactors. And on page 49, 9
I'm going to read from the abstract.
10 "In order to prevent the containment and 11 other safety-relevant components from incurring 12 serious damage caused by the detonation of the 13 hydrogen-air mixture generated during a
severe 14 accident in light-water
- reactors, passive 15 autocatalytic recombiners are used for hydrogen 16 removal.
17 "These devices make use of the fact that 18 hydrogen and oxygen react exothermically on catalytic 19 surfaces, generating steam and heat. "
20 "Experimental investigations at several 21 research facilities indicate that existing passive 22 autocatalytic recombiner systems bear the risk of 23 igniting the gaseous mixture due to an overheating of 24 the catalyst elements, caused by a strong reaction 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 35 heat generation."
1 And then, the conclusion of the paper, on 2
page 59, I want to read from that, and this more 3
addresses the issue that was addressed by Mr. Steets.
4 Actually, I'm not sure if I stated the title of Matt 5
Wald's blog entry. That's "U.S. Drops Nuclear Rule 6
Meant to Avert Hydrogen Explosions."
7 So this basically, this conclusion 8
addresses more what Mr. Steets had said. And I quote, 9
"Even if recombiners could be made safe against 10 unintended ignitions, these devices cannot solve the 11 hydrogen problem for severe accidents. Conversion 12 rates of present systems are not sufficient for 13 massive hydrogen release, and hydrogen transport to 14 the recombiners cannot be assured in a sufficient way.
15 The combination of passive autocatalytic recombiners 16 with other existing concepts for hydrogen mitigation, 17 for example, inerting or diluting, seems to be 18 advisable, even if these concepts also have 19 limitations. "
20 "One example is the reinforcement of 21 passive autocatalytic recombiners by means of 22 catalytic coated thermal insulation elements, as 23 proposed in the THINCAT project.
24 "The introduction of
- igniters, as 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 36 discussed in the past, still seems to be very 1
questionable, as the prediction of hydrogen 2
distribution and combustion in the containment is at 3
present not reliable enough to ensure safe application 4
of this measure."
5 That's the end of that quote.
6 And now, I want to move on to talking 7
about stainless steel thermal-hydraulic experiments, 8
and discuss them, because so far, I've spoken about 9
the metal-water reaction correlations, and now I want 10 to talk about heat transfer experiments that -- in the 11 petition, starting on page 104, there is a discussion 12 of a stainless steel test that's commonly included as 13 a benchmark for the validation metrics of several 14 computer codes, and this is the FLECHT-31504 test.
15 And just to briefly describe that test, 16 FLECHT-SEASET test 31504 had a rod peak power, it was 17 0.7 kilowatts per
- foot, a
reflood rate of 18 approximately one inch per second, and the PCT at the 19 onset of reflood was about 1,585 degrees Fahrenheit, 20 and the overall peak cladding temperature was about 21 2,100 degrees Fahrenheit. So, there was an increase 22 of 516 degrees Fahrenheit in that test.
23 But, I point out, in the petition, that if 24 the FLECHT-SEASET test 31504 had been conducted with a 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 37 zirconium alloy instead of with a stainless steel 1
bundle, the test results would have been very 2
different, and that with high probability, the maximum 3
temperatures of the bundle, if it were a counterpart 4
test, were done with a zirconium alloy bundle, that 5
they would have exceeded the 2,200 degrees Fahrenheit 6
peak cladding temperature limit, and the bundle would 7
- have, with high probability, incurred runaway 8
oxidation.
9 And it's very unfortunate that NRC has 10 actually never, never conducted a counterpart thermal-11 hydraulic experiment with a zirconium alloy multi-rod 12 bundle.
13 This FLECHT-SEASET 31504 test is obviously 14 very important. It's used as, like I said, a 15 benchmark for computer codes. It's discussed at 16 length in a recent study -- or program description of 17 the rod bundle heat transfer facility program that's 18 being conducted right now at Penn State. But there's 19 never been a counterpart test conducted with a 20 zirconium alloy bundle.
21 And I know I'm not supposed to ask 22 questions at the meeting, but I mean, nonetheless, I 23 mean, can you tell me why NRC has never conducted such 24 a test? And if you don't want to answer that, could 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 38 you at least think about that question?
1 It seems pretty outrageous, frankly. Here 2
you have computer codes, and the heat transfer 3
coefficients that are used in them are based off of 4
tests conducted with stainless steel bundles, and then 5
you've never conducted counterpart tests with the same 6
test parameters with zirconium alloy bundles.
7 The TH-1 tests that were conducted in the 8
early `80s, actually, one of the parameters or 9
conditions for those tests was that they didn't want 10 the peak cladding temperatures in those tests to 11 exceed 1,900 degrees Fahrenheit, and they said that 12 was actually for safety reasons.
13 And I can sympathize with people who 14 conduct tests, but the point is, the real limit is 15 2,200 degrees Fahrenheit, so NRC needs to conduct 16 tests with zirconium alloy bundles with different 17 parameters, in which the peak cladding temperature 18 would go up to at least 2,200 degrees Fahrenheit.
19 But the real problem is, if you do tests 20 like that, you're not going to have a peak cladding 21 temperature that goes up to 2,200. You're going to 22 have runaway, and it's going to just go up to the 23 point where you have rapid oxidation, and you're going 24 to have melting of the rods, etcetera.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 39 So, again, I think it's time for the NRC, 1
this experiment is obviously very important, the 2
FLECHT-SEASET 31504, to conduct a counterpart test 3
with a zirconium alloy bundle.
4 And I think that could be part of the 5
technical analysis of PRM-50-93. For example, NEI has 6
pointed out that, they say, oh, 50-93, it depends too 7
much on the results of severe fuel damage experiments, 8
and the results would be entirely different if that 9
petition were based off of the results of thermal-10 hydraulic experiments.
11 Well, the real problem I had writing that 12 is that there really is no data from thermal-hydraulic 13 experiments conducted with zirconium alloy bundles, 14 because you haven't done such experiments.
15 And in that petition, I discuss the TH-1 16 tests rather extensively. I do what I can to use that 17 very limited data to draw conclusions. I state that 18 if the initial peak cladding temperatures in those 19 tests had been higher, that they would have incurred 20 runaway oxidation.
21 But the point is, I was very limited, 22 because there's such a limited amount of data. But if 23 you really want to do an analysis of PRM-50-93, what 24 you do is you just conduct a counterpart test to 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 40 FLECHT-SEASET 31504, and you see what happens. You do 1
that with a zircaloy bundle.
2 And I just want to point out that here 3
I've been talking about tests, you know, TH-1 tests 4
from the early 1980s, and all of these FLECHT tests.
5 Those were done about 40 years ago.
6 Presently, as I discuss in the petition, 7
there is the test facility, it's the RBHT test 8
facility. That's at Penn State. And they're 9
conducting thermal-hydraulic experiments with Inconel 10 600 rods.
11 And I just want to read some information 12 that's also in the petition. Let me tell you what 13 page that's starting on. That's starting on page 116.
14 That's where I describe the current heat transfer 15 experiment program that uses Inconel 600 bundles.
16 And according to NRC's Advisory Committee 17 on Reactor Safeguards, this is a statement made in 18 2010, the rod bundle heat transfer facility program 19 was developed to address issues related to emergency 20 core cooling, including phenomena that would affect 21 peak cladding temperatures.
22 Now, as the test plan for the RBHT, the 23 test plan points out, it states, "oxidation is not 24 simulated in the RBHT facility tests since the 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 41 cladding is Inconel," and as the test plan also 1
acknowledges, "Inconel will not
- oxidize, while 2
zircaloy will oxidize and create a secondary heat 3
source at very high PCTs. Zircaloy reaction can be 4
significant at high temperatures."
5 So here you have ACRS, they're talking 6
about this program, and it's supposed to be addressing 7
issues relating to emergency core cooling, including 8
phenomena that would affect PCTs, and yet, ACRS can't 9
put two and two together? You know, that these tests 10 are conducted with Inconel 600, and as the test plan 11 explicitly states there will not be the additional 12 heat generated from the zirconium alloy that will 13 affect the PCTs? I don't know what to say. Actually, 14 I don't think I'll comment on that. I think it speaks 15 for itself.
16 So, currently, here we have the heat 17 transfer coefficients used in LOCA evaluation models, 18 they're still not based on data from thermal-hydraulic 19 experiments conducted with zirconium alloy bundles, 20 and there are currently no plans underway to make them 21 based on such data.
22 And I want to really point out that this 23
- program, the rod bundle heat transfer facility 24 program, it's not just like these tests are worthless.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 42 These tests do not help public safety. It's actually, 1
these tests decrease public safety because the results 2
of these experiments conducted with stainless steel or 3
Inconel 600 multi-rod bundles, they're going to lead 4
the interpreters of tests to false conclusions.
5 For example, a test conducted with a 6
stainless steel or Inconel 600 multi-rod bundle heated 7
up to peak cladding temperatures between 1,832 8
degrees Fahrenheit and 2,200 degrees Fahrenheit, that 9
is not going to incur runaway oxidation. It's not 10 going to have much oxidation.
11
- However, a
zirconium alloy multi-rod 12 bundle heated up to those same temperatures, between 13 1,832 and 2,200 degrees Fahrenheit, with high 14 probability, that's going to incur runaway oxidation.
15 So what we have going on, and you can read 16 this, it's in a ACRS meeting on heat transfer from 17 last year, I believe it was dated October 18th, 18 they're discussing results from this facility, the rod 19 bundle heat transfer facility.
20 And they're saying that, well, because of 21 what we see, we can actually have more power uprates.
22 We can have more power uprates, because we see that 23 our models are overly conservative. But it's a pure 24 fantasy. There's no foundation.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 43 So, like I said, all you need to do, you 1
conduct a counterpart test to FLECHT-SEASET 31504.
2 You do that with a zirconium alloy bundle and you'll 3
see -- you'll find out what's wrong with your models.
4 And I think you should just stop funding 5
programs like the RBHT program where they're not using 6
zirconium alloy bundles. And I just want to point out 7
that you're a regulator, and your job is to protect 8
public and plant workers' safety. And you need to 9
have realistic experiments.
10 And this is pretty much where I want to 11 conclude. I just want to point out that what I have 12 presented are generic issues, and what Phillip 13 Musegaas focused on I think are extremely important 14 issues that demonstrate that this petition, because of 15 where Indian Point is located, that it also addresses 16 plant-specific issues.
17 And I think it's really time that NRC has 18 Entergy lower the licensing basis peak cladding 19 temperatures of Indian Point Units 2 and 3 and also 20 has Entergy demonstrate that for Indian Point Units 2 21 and 3, that the ECCS systems would effectively quench 22 the fuel cladding in the event of LOCAs.
23 Thank you.
24 MR. MUSEGAAS: Thank you, Mark.
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 44 MR. BOSKA: This is John Boska from the 1
NRC. I did want to make one comment before we 2
continue.
3 MR. MUSEGAAS: Sure.
4 MR. BOSKA: And that is, Mr. Leyse has 5
submitted a rulemaking petition for our rule 10 CFR 6
50.46 on LOCA analysis, and his rule-making petition 7
is PRM-50-93. The NRC has accepted that petition for 8
review, and we are proceeding with the review of that 9
petition. So let me just make that one point.
10 MR. DUDLEY: On that basis, I think the 11 NRC would also like to note that we are not ignoring 12 the issues that you are raising in this phone call.
13 We are, in fact, reviewing them as part of PRM-50-93 14 and 50-95, which is actively ongoing at the present 15 time. That was Richard Dudley.
16 MR. M. LEYSE: Okay. Mark Leyse. I just 17 want to -- yes, I really appreciate that, and I am 18 aware of the fact that you're reviewing that, and I 19 appreciate that.
20 I think just more -- because it's been so 21 long that this problem has persisted, I think that's 22 more of the reason why I made some of the statements 23 that I made. But I really do appreciate that.
24 And as I had said, I really would advise 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 45 conducting counterpart tests to FLECHT-SEASET 31504.
1 I think you would really learn a lot of information 2
from that. But that's the last thing I want to say.
3 Thank you.
4 MR. MUSEGAAS: Thanks again, Mark. This 5
is Phillip Musegaas. I think that -- you know, I'll 6
hold any further comments. I'd like to hear what the 7
PRB has to say. Thank you.
8 MR. BOSKA: Thank you.
9 MR. R. LEYSE: Okay, this is Robert Leyse.
10 I just came back online. If I could just have a 11 second to talk about PRM-50-93 and the timing of its 12 review?
13 MR. M. LEYSE: Mark Leyse speaking. Of 14 course, you can.
15 MR. R. LEYSE: Well, I was asking 16 permission from the guys who run the meeting.
17 Anyway, the fact is going back to the ACRS 18 meeting of last October, October of 2010, it was at 19 that meeting that I found out that NRC had changed its 20 deadline for reviewing PRM-50-93. There's a number 21 that I don't have available, but that review was to be 22 completed by September of 2010.
23 So, Mark Leyse submitted an enforcement 24 action, and that was thrown in as a petition, and 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 46 suddenly, there was no longer any deadline. It's all 1
in the record.
2 But I never found out that the September 3
2010 deadline had gone away until Bajorek mentioned it 4
at the October 2010 ACRS meeting. And at that point, 5
I wasn't in a position to be able to talk to the ACRS.
6 So, I wouldn't want anybody to make too 7
much of the fact that PRM-50-93 is under review. I 8
- mean, apparently there was review done before 9
September of 2010, but I don't see it anywhere.
10 End of comment.
11 MR. BOSKA: All right, Mr. Leyse. Thank 12 you for your comment.
13 MR. BROWN: This is Fred Brown. Our 14 thanks to Riverkeeper for making the presentation.
15 Let me now ask the staff at NRC headquarters, do we 16 have any questions for Mr. Musegaas or Mr. Leyse about 17 the 2.206 petition?
18 No questions in headquarters?
19 The region, do you have questions?
20 MR. BICKETT: This is Brice Bickett from 21 NRC Region 1. None from Region 1.
22 MR. BROWN: Very good. Thank you.
23 At this
- time, does the licensees 24 representative or representatives have any questions?
25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 47 MR. WALPOLE: No comments or questions, 1
thank you.
2 MR. BROWN: All right. Thank you all.
3 Before I conclude the meeting, members of the public 4
may ask questions about the NRC's process for 2.206 5
petitions.
6 However, as stated at the opening, the 7
purpose of this meeting is to provide an opportunity 8
for the public to question or examine -- is NOT an 9
opportunity for the public to question or examine the 10 PRB regarding the merit of the petition request.
11 Are there any questions?
12 (Pause.)
13 Hearing
- none, I
want to thank the 14 petitioners for their time and energy to make their 15 presentation to the Petition Review Board and to 16 provide the staff with clarifying information as to 17 the petition that you have submitted.
18 Before we close -- we don't have a court 19 reporter online?
20 COURT REPORTER: Yes, we do. I'm online.
21 I do have two questions.
22 MR. BROWN: Thank you, please.
23 COURT REPORTER: The first question is if 24 the people online from Entergy could identify 25
NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.
(202) 234-4433 WASHINGTON, D.C. 20005-3701 www.nealrgross.com 48 themselves. I didn't catch them the first time they 1
identified themselves.
2 MR. WALPOLE: Sure, it's Bob Walpole from 3
Indian Point.
4 COURT REPORTER: W-A-L-P-O-L-E?
5 MR. WALPOLE: That is correct.
6 COURT REPORTER: And there was one other 7
person from Entergy.
8 MR. WALPOLE: His name was Adi Irani, I-R-9 A-N-I, first name A-D-I, Alpha Delta Indigo.
10 COURT REPORTER: Thank you very much. The 11 second question is if there is someone from the NRC 12 staff who I could call if I have questions. I don't 13 anticipate any now, but as I go over my notes, I 14 might. Is there someone who can give me their phone 15 number?
16 MR. BOSKA: You should call John Boska at 17 301-415-2901.
18 COURT REPORTER: 2901. Thank you very 19 much. It will be a call within the hour or none at 20 all.
21 MR. BOSKA: Thank you.
22 MR. BROWN: Thank you all.
23 (Whereupon, the above-entitled matter was 24 concluded at 2:15 p.m.)
25