ML17059C395

Transcript of 10 CFR 2.206 Petition Review Board Conference Call Re Braidwood and Byron, February 1, 2017, Pages 1-73

ML17059C395
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Issue date:	02/01/2017
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NRC-2834
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Official Transcript of Proceedings NUCLEAR REGULATORY COMMISSION

Title:

Docket Number:

Location:

Date:

Work Order No.:

10 CFR 2.206 Petition Review Board (PRB)

Conference Call RE Braidwood/Byron (n/a) teleconference Wednesday, February 1, 2017 NRC-2834

!oRIGINALI NEAL R. GROSS AND CO., INC.

Court Reporters and Transcribers 1323 Rhode Island Avenue, N.W.

Washington, D.C. 20005 (202) 234-4433 Pages 1-73

1 1

UNITED STATES OF AMERICA 2

NUCLEAR REGULATORY COMMISSION 3

+ + + + +

4 10 CFR 2.206 PETITION REVIEW BOARD (PRB) 5 CONFERENCE CALL 6

RE 7

2834 BRAIDWOOD/BYRON 8

+ + + + +

9 WEDNESDAY

'\\

10 FEBRUARY 1, 2017 11

+ + + + +

12 The conference call was held at 1: 00 p. m.,

13 Michael Case, Chairperson of the Pe ti ti on Review 14 Board, presiding.

15 16 PETITIONER: SAMUEL MIRANDA 17 18 PETITION REVIEW BOARD MEMBERS 19 MICHAEL CASE, PRB Chairman, Office of Nuclear 20 Regulatory Research 21 MERRILEE BANIC, 2.206 Petition Coordinator, 22 Office of Nuclear Reactor Regulation 23 ROBERT BEATON, Alternate PRB Member, Office of 24 Nuclear Reactor Regulation 25 JOHN BILLERBECK, PRB Member, Office of Nuclear (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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Reactor Regulation 2

JOSHUA BORROMEO, Alternate PRB Member, Office 3

of Nuclear Reactor Regulation 4

TIMOTHY DRZEWIECKI, PRB Member, Office of New 5

Reactors 6

GLADYS FIGUEROA-TOLEDO, PRB Member, Office of 7

Enforcement 8

SARA KIRKWOOD, PRB Member, Office of General 9

Counsel 10 JOEL WIEBE, PRB Petition Manager, Office of 11 Nuclear Reactor Regulation 12 13 NRC HEADQUARTERS STAFF 14 RUSSELL ARRIGHI, Office of Enforcement 15 ERIC DUNCAN, Region III 16 JAMES McGHEE, Region III 17 ERIC OESTERLE, Office of Nuclear Reactor 18 Regulation 19 20 PUBLIC COMMENTERS 21 MARVIN LEWIS 22 23 24 25 NEAL R. GROSS (202) 234-4433 COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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P R 0 C E E D I N G S 2

(1:00 p.m.)

3 MR.

WIEBE:

The purpose of this public 4

meeting and teleconference is for the Pe ti ti oner, 5

Samuel Miranda to address the Petition Review Board 6

for the petition he submitted dated November 15, 2016, 7

for Braidwood and Byron.

8 First of all my name is Joel Wiebe.

I am 9

the NRC petition manager for this petition.

We are 10 here today to allow the petitioner to address the 11 Petition Review Board regarding his 2.206 petition.

12 As part of the PRB' s review of this 13 petition, Mr. Miranda has requested this opportunity 14 to address the PRB.

The meeting is scheduled from 15 1:00 to 3:00 p.m.

16 The meeting is being recorded by the NRC 17 Operations Center and will be transcribed by a court 18 reporter.

The transcript will become a supplement to 19 the petition.

This transcript will also be made 20 publicly available.

21 I'd like to open this meeting with 22 introductions.

The PRB Chair is Michael Case.

23 Michael is the Director of the Division of Engineering 24 in the Office of Nuclear Regulatory Research.

25 (202) 234-4433 I'd like the rest of your to, the rest of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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the Pe ti ti on Review Board to introduce themselves.

As 2

we go around the room please be sure to clearly state 3

your name, your position and the office that you work 4

for within the NRC for the record.

5 MS.

FIGUEROA-TOLEDO:

Good afternoon, 6

everyone.

My name is Gladys Figueroa.

I work in OE 7

and I'm an enforcement specialist.

8 MR.

CASE:

Okay.

I'll do myself again.

9 This is Mike Case.

I'm the, really I'm the director 10 of the Division of Safety Analysis and Research.

I 11 used to be the Division of Engineering in the Office 12

• of Research.

13 MR.

BEATON:

My name is Robert Beaton.

14 I'm a technical reviewer in the Reactor Systems Branch 15 in NRR.

16 MR. BORROMEO:

My name is Josh Borromeo in 1 7 NRR and I'm a technical reviewer in the Reactor 18 Systems Branch.

19 MR. OESTERLE: Eric Oesterle, Chief of the 20 Reactor Systems Branch in the Office of Nuclear 21 Reactor Regulation.

22 MS.

BANIC:

Ms.

Banic, Petition 23 Coordinator, NRR.

24 MR. MIRANDA:

Samuel Miranda, Petitioner.

25 (202) 234-4433 MS.

KIRKWOOD:

Sara Kirkwood, Off ice of NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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General Counsel.

2 PARTICIPANT:

(Telephonic interference),

3 Office of the Inspector General.

4 MR. BILLERBECK:

I'm John Billerbeck. I'm 5

a mechanical engineer in the Division of Engineering.

6 MS. GONZALEZ:

I'm Dee Gonzalez and I'm 7

with the Office of the Inspector General.

8 MR.

WIEBE:

We've completed the 9

introductions of the NRC staff at the NRC 10 headquarters.

Are there any NRC petitioners from 11 headquarters on the phone?

12 MR.

ARRIGHI:

This is Russell Arrighi.

13 I'm from the Office of Enforcement.

14 MR. WIEBE:

Joel, Eric Duncan is on the 15 line, the branch chief for Byron and Braidwood here in 16 Region III in Lyle.

Okay.

Are there any other NRC 17 participants from Region III?

18 MR. MCGHEE:

Joel, James McGhee from Byron 19 as the senior resident inspector.

20 MR. WIEBE:

Okay.

Is the court reporter 21 on the line?

22 COURT REPORTER:

Yes, sir.

23 MR. WIEBE:

Okay.

Thank you.

All right.

24 If there are any licensee personnel on the line I 25 would like each of you to email me your name, position (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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and organization.

I would also like members of the 2

public to do the same if those members of the public 3

wish to be identified as attending the meeting.

4 You're not required to introduce yourself 5

as a member of the public.

But if you do wish to do 6

so please email me.

My email is joel.wiebe@nrc.gov.

7 Okay.

Sam, you already introduced yourself.

So I 8

won't have you do it again.

9 I'd like to emphasize that we need to 10 speak clearly and loudly to make sure that the court 11 reporter can accurately transcribe the meeting.

If 12 you do have something to say please state your name 13 for the record first.

14 For those who are dialing into the meeting 15 please remember to mute your phones to minimize any 16 background noise.

If you do not have a mute button 17 you can mute the phone by pressing the key star and 18 then 6.

And to unmute you would press star and 6 19 again.

20 At this time I'll turn it over to the PRB 21 Chairman.

Michael.

22 MR. CASE:

Thank you.

Welcome, everybody.

2 3 This meeting is in regards to a

2. 2 0 6 petition 24 25 submitted by Sam Miranda.

I'd like to share some background on our process first.

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And most of you probably know that.

2 You've probably been around and heard this.

Section 3

2.206 of Title 10 of the Code of Federal Regulations 4

describes the petition process and it's the primary 5

mechanism for the public to request enforcement action 6

by the NRC in a public way.

7 This process permits anyone to petition 8

the NRC to take enforcement type actions related to 9

NRC licensees or licensed activities.

Depending on 10 the results of the evaluation the NRC could modify, 11 suspend or revoke the NRC issued license or take any 12 other appropriate enforcement action to resolve the 13 problem.

14 The NRC staff guidance for the disposition 15 of 2. 2 0 6 petitions requests is in Management Directive 16 8.11 which is publicly available and it's pretty fat, 17 good reading.

The purpose of today's meeting is to 18 give the petitioner an opportunity to provide

• any 19 additional explanation or support for the petition 20 before the Petition Review Board initial 21 considerations and recommendations.

22 And of course that makes a lot of sense 23 because, you know, I've read the petition about three 24 or four times now.

And so it's always great to really 25 hear it from the first person because it gives a lot (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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of good insights.

2 The purpose of today's meeting is to give 3

Sam the opportunity to do that for us.

And the 4

meeting is not a hearing nor is it an opportunity for 5

the petitioner to question or examine the PRB on the 6

merits of the issues presented in the request.

7 No decisions regarding the merit of the 8

petition will be made at this meeting.

And then 9

following the meeting the Petition Review Board will 10 conduct its internal deliberations.

11 The outcomes of this internal meeting will 12 be discussed with the petitioner. The Petition Review 13 Board consists of me, the Chairman, usually a manager 14 in the SES core at the NRC.

It has a petition manager 15 and a Petition Review Board 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 so we get experts 19 to help us.

And the members have already introduced 20 themselves.

21 As described in our process the NRC staff 22 may ask clarifying questions in order to better 23 understand the petitioner's presentation and to reach 24 a reasoned decision whether to accept or reject the 25 petitioner's request for review under the 2.206 (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1.323 RHODE ISLAND AVE., N.W.

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process.

2 Also described in our

process, the 3

licensee has been invited to participate in today's 4

meeting to ensure that it understands the concerns 5

about their facilities or activities.

The licensees 6

may ask questions to clarify issues raised by the 7

petitioner.

8 I want to stress that the licensees are 9

not part of the PRB decision-making process.

And 10 licensees will have an opportunity to ask petitioners 11 questions after his presentation.

12 I'd like to *summarize the scope of the 13 petition under consideration and the NRC activities to 14 date.

The petitioner identified eight omitted points 15 and 11 mistaken points regarding the licensee's 16 commitments to NRC guidance and related standards 17 associated with the inadvertent operation of an 18 emergency cooling system during power operation event 19 and the non-escalation guidance.

20 The petition also states that a

new 21 accident is created without addressing it in the most 22 significant hazard statement associated with power 23 uprate amendment.

As a

reminder for our phone 24 participants, please identify yourself if you make any 25 remarks as this will help us in the preparation of the (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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meeting transcript that will be made publicly 2

available.

3 Since this is a public meeting I would 4

like to remind the PRB Members, the licensees, the 5

petitioners and any other participants of the need to 6

refrain from discussing any sensitive NRC or 7

proprietary information during today's public meeting.

8 Okay, Mr. Miranda, I will turn it over to 9

you now and allow you the opportunity to provide any 10 information you think the PRB should consider as part 11 of this petition.

And thanks for coming.

12 MR.

MIRANDA:

Thank you.

My name is 13 Samuel Miranda.

I submitted the Enforcement Petition 14 that is under your review.

It's submitted under the 15 terms of 10 CFR Section 2. 206 and concerns the 16 licensing and operation of the Byron and Braidwood 17 Stations.

18 I maintain that Exelon, by the way I'm not 19 going to read you the petition.

You've read it.

You 20 know what's in it.

I'm going to supplement it here.

21 I maintain that Exelon has obtained the 22 NRC' s authorization to operate its Byron and Braidwood 23 Stations at an uprated power level without first 24 demonstrating that its plant designs meet all design 25 requirements specified in the licensing basis.

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I'm reque~ting among other things that the 2

NRC revoke its uprating approvals and compel Exelon to 3

operate its Byron and Braidwood Stations at their 4

originally licensed power levels until the required 5

compliance is made.

6 The petition describes several issues in 7

Exelon' s

LAR, License Amendment Request and it's 8

licensing basis that support this request.

I will not 9

read you the petition.

It's on record and available 10 for discussion at any time you choose.

11 Instead, I will use my allotted time to 12 augment and elaborate upon the petition.

I will also 13 have a couple of handouts for your reference and the 14 record.

15 By the way, non-escalation is more than a 16 guidance.

It's a design requirement.

Before I get 17 into the details of these issues I will provide a 18 brief but relevant background concerning my education 19 and experience.

I will also list some disclosures 20 that pertain to the issue.

21 If you have any questions I ask that you 22 hold them until the end of my presentation.

I don't 23 want to run over my time.

I earned Bachelor's and 24 Master's degrees in Nuclear Engineering from Columbia 25 University and hold a professional engineers license (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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in mechanical engineering in the Commonweal th of 2

Pennsylvania.

3 I have more than 40 years of experience in 4

reactor safety analysis and licensing at Westinghouse 5

and the NRC.

I worked 25 years at Westinghouse in 6

their Nuclear Safety Department where I performed 7

accident analyses of the kind that are under review 8

today.

9 I also developed standards and methods for 10 use in nuclear safety analysis and automatic reactor 11 protection systems design which included the 12 preparation systems functional requirements, component 13 sizing and determination of setpoints, time response 14 limits and technical specification revisions.

15 In the 1980s I managed a program for the 16 Westinghouse Owners Group to reduce the frequency of 17 unnecessary reactor trips.

This was known as the WOG 18 TRAP, W-0-G, Westinghouse Owners Group, T-R-A-P, Trip 19 Reduction Assessment Program.

20 Like unnecessary reactor trips, serious 21 safety injection actuations are very frequent.

22 Sometimes safety injection actuations accompany 23 unnecessary reactor trips.

24 25 After Westinghouse I

worked as a

contractor at the Salem Nuclear Plant where I compared (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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the LAR for qualifying Salem's pressurizer power 2

operated relief valves of the PORVs for water relief, 3

improving their automatic control system circuitry, 4

revising plant tech specs and meetings all the other 5

requirements for upgrading the valves to satisfy 6

safety grade status.

7 The NRC approved that LAR in 1997 making 8

Salem the first plant with safety grade water 9

qualified power operator relief valves.

I worked for 10 14 years at the NRC in the NRR's Division of Safety 11 Systems.

12 I've revised several sections of the 13 standard review plan, presented those revisions to the 14 ACRS and I wrote RIS 2005-29 regarding compliance with 15 the design requirement that is the subject of this 16 petition.

I retired from the NRC in August 2014 at 17 Grade Level 15.

18 I hereby make the following disclosures.

19 I was directly involved, as an NRC employee I was 20 directly involved in the imposition of License 21 Condition 2. K for the Seabrook Plant in 2005 regarding 22 compliance with the non-escalation requirement.

23 I wrote RIS 2005-29 and the first draft of 24 25 RIS 2005-29 Revision 1.

Have reviewed the power uprating LARs regarding the Byron and Braidwood plants (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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in 2014 and at that time I withheld my concurrence.

2 I would also add to this disclosure that 3

at no time while I was working for Westinghouse did I 4

have involvement in producing any of the Nuclear 5

Safety Advisory Letters that we will be discussing 6

later.

The petition concerns all PWRs, particularly 7

the Byron and Braidwood plants which are currently 8

operating in Byron and Braceville, Illinois.

9 Byron 1 was the first plant licensed in 10

'85 and then Byron 2 in '87.

Braidwood 1 in '88 and 11 Braidwood 2 also in I 88.

Together these plants have 12 been operating for about 118 years, 118 reactor 13 operating years.

14 We have about ten years to go in 15 operation.

This is a, no, they have 38 years to go.

16 It's an average of about ten years per plant.

Other 17 PWRs that are operated by Exelon are Arkansas Nuclear 18 1 and 2, Calvert Cliffs 1 and 2 and Three Mile Island 19 Unit 1 and R.E. Ginna.

20 R. E.

Ginna is a special interest here 21 because it's a two loop Westinghouse plant and I would 22 point out that in 1974 a plant in Switzerland, Beznau 23 had an occurrence wherein they took one of their 24 turbines, it's a two turbine plant.

So it amounts to 25 a 50 percent power reduction.

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This is, a 50 percent power reduction is 2

something that a Westinghouse plant is designed to 3

tolerate.

It should stay on line without a reactor 4

trip.

5 Well nevertheless this plant tripped and 6

pressurized in the reactor coolant system to the point 7

of opening their PORVs.

And when the time came for 8

the PORVs to reseat one of the PORVs stuck open, 1974 9

August.

10 Westinghouse sent a

couple of its 11 engineers from the Brussels office to investigate and*

12 they found that this PORV that was stuck open was 13 stuck open for a good reason.

It was broken.

The 14 valve broke, was broken.

I think there's a cast iron 15 design.

16 Westinghouse did not report this to the 17 NRC at the time, did not report this to anyone.

Then 18 in 1979 after Three Mile Island occurred in which they 19 had a stuck open PORV this information was disclosed 20 and the NRC took great interest in this.

21 After all, it was a foreign plant that the 22 NRC was concerned anyway because that PORV at the 23 Beznau unit was exactly the same PORV, same design, 24 same materials as used in our; in the Ginna plant, 25 stuff that Ginna should have known, also in the PORV (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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unit too.

2 The NRC at that time,

'79 issued an 3

information notice, Notice Number 7 9-4 5 where they 4

described what happened at Beznau.

They included it 5

actually in the internal Westinghouse report of the 6

investigation of that incident to inform the American 7

licensees.

8 This petition concerns three events that 9

are analyzed and reported in Chapter 15 of the Byron 10 and Braidwood FSAR.

The inadvertent operation of 11 emergency core cooling system during power operation, 12 also called IOECCS.

13 The chemical and volume control system 14 malfunctioned, which by the way I did not cover in the 15 petition but it's one of the events.

And the 16 inadvertent opening of the pressurized relief or 17 safety valve.

This is also not covered in the 18 petition but the same principles apply.

19 These refer to categorized as anticipated 20 operation occurrences or AOOs.

AOOs are defined in 21 the general design criteria as, and I'm quoting, 22 "those conditions of normal operation which are 23 expected to occur one or more times during the life of 24 the nuclear power unit".

25 (202) 234-4433 In its FSAR Exelon commits to meet the NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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following requirements for AOOs and there are three 2

requirements.

One, pressure in the reactor coolant 3

and main steam system should be maintained below 110 4

percent of* design values.

5

Two, fuel cladding integrity shall be 6

maintained by ensuring that the minimum departure from 7

nuclear boiling ratio, DNBR remains above the DNBR 8

limit derived at a 95 percent confidence level and 95 9

percent probability.

And the criterion of interest 10 here an incident of moderate frequency should not 11 generate a more serious plant condition without other 12 faults occurring independently.

13 Exelon also committed to meet another 14 requirement which was specified by the American 15 Nuclear Society in 1973. It states that AOOs shall be 16 accommodated with at most a shutdown of the reactor 1 7 coolant, a shutdown of the reactor with the plant 18 capable of returning to operation after corrective 19 action.

20 So in this case these events that I

21 mentioned earlier the first two requirements are easy 22 to meet.

The charging pumps in the Byron and 23 Braidwood plants, as well as other PWRs, simply cannot 24 pressurize a reactor coolant system 110 percent of its 25 design value.

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When actuated during an inadvertent ECCS 2

actuation or when actuated in an ECCS actuation they 3

pressurized the reactor coolant system to their shut 4

off head and no more.

So analysis and their shut off 5

head, by the way, is about 2,600 psi.

6 But they' re never going to get to 110 7

percent of reactor coolant system design pressure 8

which is 2,750 psi.

So no analysis is necessary to 9

demonstrate compliance with this first requirement.

10 Nevertheless, Exelon has one in Chapter 11

15.

The reason I mention this will become evident 12 later.

The second requirement specified that an AOO 13 may not breach the fuel cladding integrity when the 14 ECCS is actuated the reactor is immediately tripped as 15 part of the ECCS actuation sequence.

16 First thing, the reactor is tripped. It's 17 hard to imagine a threat to cladding integrity when 18 the reactor is not generating power.

Still no IOECCS 19 analysis is necessary to demonstrate compliance with 20 the second requirement. Nevertheless, Exelon provides

21.

one.

22 Predictably, Exelon' s analysis results, 23 and you can find these in Chapter 15, show there is no 24 approach to DNB at any time during the event.

25 However, the CVCS malfunction does not lead to an (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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immediate reactor trip.

2 An analysis is necessary to demonstrate 3

that the reactor is automatically tripped before any 4

fuel damage can be incurred.

Exelon does not provide 5

an analysis.

6

Instead, Exelon points to another 7

dissimilar event analysis.

The Byron and Braidwood 8

FSAR Chapter 15 states, and this is in Section 15.5.2, 9

chemical and volume control system malfunction that 10 increases reactor coolant inventory.

11 And this is what Exelon says. An increase 12 in reactor coolant inventory which results from the 13 addition of cold, unborated water to the reactor 14 coolant system is analyzed in Subsection 15.4.6, 15 chemical and volume control system malfunction that 16 results in a decrease of boron concentration in the 17 reactor coolant.

18 An increase in reactor coolant inventory 19 which results from the injection of highly borated 20 water into the reactor coolant system, that would be 21 this event the inadvertent ECCS actuation, is analyzed 22 in Subsection 15. 5. 1. Inadvertent operation emergency 23 core cooling operation during power, inadvertent 24 operation, emergency, in other words what they' re 25 doing is they're pointing to a different analysis.

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The eves malfunction that decreases boron 2

concentration is a reactivity anomaly, not a mass 3

addition event.

This distinction is made by the Reg.

4 Guide 1. 70.

It cannot be used to address a mass 5

addition event.

6 So at this point I would like to add this 7

to the 11 errors that were in the petition.

This is 8

error number 12.

So and for those of you familiar 9

with the two analyses involved one increases the mass 10 of the water inventory in the reactor coolant system 11 causing it to pressurize and fill the pressurizer, et 12 cetera, et cetera.

13 The other one causes a

reactivity 14 excursion.

The two analyses are very different.

One 15 involves a computer simulation of the plant.

The 16 other one is basically a hand calculation which 17 balances reactivity in the core.

18 It's nothing to do with the pressurizer or 19 the pressurizer water level.

So Exelon does not 20 provide an analysis for the IOEees for the inadvertent 21 opening of a pressurized relief safety valve to 22 demonstrate this.

This is common.

Other plants 23 don't provide that either to demonstrate compliance 24 25 with the third requirement, the non-escalation requirement.

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Therefore, the Byron and Braidwood 2

licensing basis as you're quoted in the FSAR contains 3

two unnecessary analyses which indicate, they don't 4

know what their analyses are all about and why they' re 5

done and lacks three analyses.

6 The missing analyses are necessary, the 7

three missing analyses are the ones that are necessary 8

to demonstrate compliance with the non-escalation 9

requirement.

Therefore, Exelon has not demonstrated 10 that its Byron and Braidwood plants comply with the 11 non-escalation requirement.

12 And these plants have been operating now 13 for almost 118 years.

So why is it important to 14 demonstrate compliance with the non-escalation 15 requirement?

If risk is defined as the product of 16 consequences and the frequency of occurrence then the 17 risk of an AOO would be about the same as the risk of 18 a LOCA.

19 This principle is at the core of nuclear 20 plant design and licensing.

In 1983 the AMS, American 21 Nuclear Society stated the nuclear safety criteria 22 have been established on the premise that a) those 23 situations in the plant that are assessed as having a 24 high frequency of occurring shall have a

small 25 consequence to the public.

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And b) those extreme situations having the 2

potential for the greatest consequence to the public 3

shall be those having a

very low frequency of 4

occurrence.

I call it the constant risk principle.

5 Plus, don't comply with the non-escalation requirement 6

it is possible to create high frequency, high 7

consequence events.

8 I maintain that this is the situation at 9

the Byron and Braidwood plants since Exelon has not 10 demonstrated that its plants comply with the non-11 escalation requirement.

12 Exelon' s compliance rationale which can be 13 found in Chapter 15. 5. 1 in the FSAR, inadvertent 14 operation of ECCS and also in its applications for two 15 power upratings subsequently claims that the non-16 escalation requirement is met by qualifying its 17 pressurizer safety valves, which I could also refer to 18 as PSVs, the water relief and operating them in lieu 19 of its power operated relief valve.

20 Exelon states, "the SI flow results in 21 liquid discharge through the pressurizer safety relief 22 23 24 25 valves."

In order for the PSVs to open the PO RVs would have to remain closed.

This would not be a conservative assumption and analyses that are performed to show compliance with the non-escalation (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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requirement.

This is an assumption one might use to 2

show compliance with that over pressure limit, not the 3

non-escalation requirement.

4 As every safety analyst here in the room 5

will tell you, the same analysis can be done in many 6

different ways depending upon what the, what you 7

intend to prove.

A conservative assumption in one may 8

be non-conservative in another.

9 This assumption prevents the possibility 10 of a PORV failure.

You assumed, the thing you're 11 looking for you're just assuming it away.

Since the 12 licensing basis does not contain an IOECCS analysis to 13 show compliance one must construe this to mean either 14 that one, you' re looking at an overpressure case 15 analysis in which case it would be conservative or 16 two, the PORVs have to be kept closed.

17 Somehow they're going to be kept closed.

18 Does Exelon intend to operate with isolated PORVs or 19 does Exelon intend to instruct its operators to close 20 the PORVs as soon as they see the pressurizer pressure 21 rising?

22 Neither of these possibilities, in my 23 opinion, is acceptable.

The PORVs not the PSV were 24 designed to operate during AOOs.

The PORVs as well as 25 pressurizer spray and heaters comprise the pressurizer (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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pressure control system.

2 They're designed to prevent unnecessary 3

reactor trips and challenges to the PSVs.

By the way, 4

you can see this challenge is to the PSV language in 5

a number of places in licensing basis analyses and 6

also in the Westinghouse Nuclear Safety Advisory 7

letters.

8 The PORVs are designed to relieve enough 9

pressure to keep the plant online during AOOs, for 10

example, during turbine trips or partial load 11 rejections.

Exelon does not explain how the PSVs 12 which are intended for accidents that are not expected 13 to occur more than once in a plant's lifetime could 14 reasonably be expected to open and reseat as often as 15 several times a year.

16 At this time I would like to hand out a 17 graph on an, I have 15 here.

So this last part can be 18 seen in this little picture I have here, Figure 1.

So 19 the, I call this is AOO boundary.

20 Recall that requirement that I mentioned 21 earlier it comes from the American Nuclear Society 22

Standard, ANS

18. 2-1973 which just about all PWR 23 operators commit to meeting.

24 That standard says AOO shall be 25 accommodated with, at most, a shutdown of the reactor (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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with the plant capable of returning to operation after 2

corrective action.

I'm not going to worry about that 3

second part of that.

4 Capable of returning to operation after 5

corrective action could be construed in a number of 6

different ways and subject to discussion and argument 7

that could take, you know, days.

I'm just going to 8

talk about shutting down the reactor.

9 If you look at this little plot here that 10 red line the 2,400 PSI is where the reactor trips on 11 high pressurizer pressure.

So what that means is 12 everything to the left of the line would be an AOO and 13 everything to the right of the line would not be an 14 AOO.

15 So if it's not an AOO what is it? It's a 16 Condition 304 event. Those are the only possibilities 17 left.

So what happens is that little blue line at the 18 bottom that is the steam relief rate of one PORV.

19 That opens at 2,350 PSI and this is enough to prevent 20 the opening of any safety valves.

21 This is during an AOO.

If what you have 22 is an AOO and nothing more serious than one PORV 23 opening and relieving steam should be enough to limit 24 the pressure to 2,350 PSI which would avoid number 25 one, a reactor trip and number two, opening of any (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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safety valves.

2 The plant basically, the plant is still 3

online. This figure also shows that little green line 4

is the opening of three safety valves and relieving 5

steam.

And they' re opening setpoint is 2, 500 PSI 6

which puts them by definition outside the realm of an 7

AOO.

8 They don't open until after you have 9

something much more serious than an AOO.

And I have 10 three safety valves here because if you have reached 11 2, 500 PSI that is the condition for opening of the 12 safety valves.

13 It could be, if one opens all three will 14 open.

So what I'm saying, I titled this plot one PORV 15 or three safety valves.

That is what happens during 16 an AOO or during any, actually during any, if you have 17 one PORV open there are no safety valves open.

18 If you have a situation where, if you have 19 a situation where you need relief through all of the 20 PORVs and all of the safety valves it's not an AOO and 21 chances are it's not even a Condition 3 or 4 event.

22 It could be beyond design basis event like an ATWS.

23 24 25 In the analyses, in fact I wrote the subject capsule and submitted it.

I also wrote the last Westinghouse report which was submitted in 1979 (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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before the adverse rule was promulgated.

And those 2

analyses for the loss of feedwater and the loss of 3

load after analyses it showed that all the PORVs and 4

all the safety valves opened would need the other and 5

still the pressure reaches levels greater than 3,000.

6 So if you're opening safety valves you're 7

not mitigating an AOO.

It's already escalated beyond 8

the AOO.

You've already violated the non-escalation 9

criterion.

10 So it has to be a Condition 3 event, maybe 11 a Condition 4 event.

Worse, it's a Condition 3 event 12 with the frequency of occurrence of an AOO because 13 it's initiated by an AOO.

14 Worse still the frequency of occurrence 15 will be the sum of the frequencies of occurrence of 16 all AOOs that pressurize the RCS with the opening set 17 pressure of the PORVs.

So now we're including the, 18 besides the IOECCS and the CVCS malfunction, we're 19 also including things like loss of feedwater, loss of 20 load, turbine trip.

21 Exelon's compliance strategy which 22 prevents the PORVs from opening allows the RCS 23 pressure to exceed by some hundred PSI the reactor 2 4 trip set point in order to open the PSVs, you' re 25 relying on the PSVs.

This pressure level is beyond (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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the defined grade of an AOO.

2 So therefore, it becomes necessary, in 3

this compliance strategy it becomes necessary to 4

generate a more serious plant condition in order to 5

open any of the PSVs.

In this respect Exelon begs the 6

question, in other words, Exelon claims that certain 7

ANS Condition 2 events must be allowed to progress to 8

more serious ANS Condition 3 events in order to 9

demonstrate that those ANS 2 events will not progress 10 to the most serious ANS Condition 3 events.

11 Does that make sense? It shouldn't. It's 12 like Vietnam when the American troops were destroying 13 villages in order to save them.

14 MR. CASE:

Robert doesn't remember that, 15 but I do.

16 MR.

MIRANDA:

There's also an invalid 1 7 comparison between two similar events.

This is in the 18 petition.

I'll just mention it.

19 Exelon claims that if the pressurizer, 2 0 this is a quote, "if the pressurizer safety relief 21 valves do not reseat then the transient will proceed 22 and terminate as described in Section 15.6.1 23 inadvertent opening of a pressurizer safety or relief 24 valve."

25 This event is also caused by, as an event (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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of moderate frequency.

An event of moderate frequency 2

is also known as an AOO or a Condition 2 event.

This 3

is not true.

They're not the same.

4 There isn't even a

common basis for 5

comparison.

And I' 11 just give you a few of the 6

differences. If we look at the inadvertent opening of 7

a pressurizer safety valve which is in the FSAR, for 8

some reason, by the way, there is no such thing as an 9

inadvertent opening of a pressurizer safety valve.

10 There is no control system.

There is no 11 reason for it to open other than high pressure.

It's 12 a spring loaded valve.

If you have the pressure to, 13 if you, if it opens at any other time it's due to a 14 mechanical fault like a broken spring.

15 So that is beyond the frequency of 16 occurrence that would be considered for an AOO.

It's 17 already a Condi ti on 3 event.

But it's analyzed, 18 nevertheless, in Chapter 15 as a Condition 2 event 19 because it's conservative.

20 The pressurizer safety valve *is twice the 21 size of a PORV.

A PORV can open spuriously.

They 22 covered that.

They said we' 11 just analyze the bigger 23 valve and the result therefore is going to be 24 constrainable.

25 There are problems with that which we (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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don't need to get into here because I don't think we 2

have the time.

Anyway, if you spuriously open a PSV 3

well that's going to occur at 2, 250 PSI.

That's 4

normal operating pressure of the reactor, 2,250 PSI.

5 On the other hand, if you have an IOECCS 6

that proceeded into the opening of a safety valve, 7

okay, and it's stuck open you pass water through the 8

safety valve and it's stuck open well that occurs at 9

2,500 PSI.

And if you want to look at a spurious 10 opening of a PORV well that's still 2,350 PSI.

11 Either way it's more than the 2,250 PSI 12 that is analyzed from the Chapter 15.

Furthermore, 13 the analysis of this event as an AOO is concerned with 14 meeting that section criteria of reserving fuel 15 cladding integrity.

16 They' re looking at whether or not the 17 depressurization that results from opening one of 18 these valves is going to be detected either by the low 19 pressurizer pressure reactor trip or more likely 20 actually the source of the reactor trip should be the 21 lone thermal margin you see in the plant, or the over 22 temperature delta in the Westinghouse plant it will 23 detect a degradation of thermal monitoring and trip 24 the reactor.

25 (202) 234-4433 Well if you look at Chapter 15 analyses it NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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doesn't take long to get there.

This is an analysis 2

that lasts about five seconds.

The reactor trip 3

occurs, the analysis is over.

You've proven your 4

point.

5 No fuel damage.

And so ECCS is never even 6

actuated.

But here is Exelon telling us that if we 7

have an inadvertent operation of an ECCS and we open 8

the safety valve and pass water through it and it 9

sticks open well just go to Section 15. 6. 1 and look at 10 this analysis of the inadvertent opening of a safety 11 valve.

12 It's a

dead end.

It's an invalid 13 comparison on many different levels.

I would also add 14 that if you do stick open a safety valve, okay, as 15 Exelon suggests and say well we're going to pass water 16 through the safety valve.

They're going to reseat.

17 In that case if one is open just look at 18 this other analysis in Chapter 15 of the opening of 19 the safety valve.

If it does, if a safety valve does 20 stick open, how do you know?

How do you know it's 21 stuck open?

22 You don't know this until the pressurizer 23 pressure drops to a

level where you expect the 24 25 pressurizer safety valve to close, okay.

If it's stuck open it's going to be depressurizing the RCS and (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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32 l

it should close.

2 Meanwhile, you know you have an 3

inadvertent ECCS actuation.

So what are you doing?

4 You're trying to shut off the ECCS, the event that's 5

causing the pressurization.

6 So if while you're shutting off the ECCS 7

this valve sticks open you expect it to close and it 8

doesn't meanwhile you've succeeded in closing, in 9

shutting off the ECCS PORV, now you have something 10 that looks a lot like Three Mile Island.

Three Mile 11 Island had a stuck open PORV with no ECCS.

12 This is worse.

This is a stuck open 13 safety valve with no ECCS.

There's no possibility of 14 isolating a safety valve.

So the only option the 15 operators have at that point would be to somehow get 16 the ECCS started again.

17 And three open safety valves, which you 18 could have, if you stick open one you can stick open 19 three, if a person sticks open three safety valves now 20 you have a

break at the top of the pressurizer 21 equivalent to a

3.7 inch hole.

That would be 22 basically a 3.7 inch pipeline break.

23 This, I also mention this in the petition.

24 ECCS will not match pressurizer safety valve water 25 relief rate.

Exelon claims that the flow through a (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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stuck open safety valve would be a minor RCS leak, RCS 2

is reactor coolant system.

3 It states and this is Exelon, this is from 4

the Exelon FSAR, American Nuclear Society Standard 5

51.1/NlS. 2-1973 describes a Condition 2 event as a 6

minor reactor coolant system leak which would not 7

prevent orderly shutdown and cool down assuming makeup 8

is provided by normal make up systems only.

9 Normal make up systems are defined as 10 those systems normally used to maintain reactor 11 coolant inventory under respective conditions of start 12 up, hot stand by, power operation or cool down using 13 on site power.

If the cause of the water relief is 14 the ECCS flow, I'm still quoting here, "the magnitude 15 of the leak will be less than or equivalent to that of 16 the ECCS."

17 That is operation of the ECCS maintains 18 RCS inventory during the postulated event and 19 establishes a magnitude of the subject.

This, by the 20 way, is copied directly from a Westinghouse Nuclear 21 Safety Advisory letter NSAL 93-013 issued in I think, 22 it was issued in '93.

23 The situation that Exelon describes 24 wherein the flow through the open PSV would be matched 25 by the flow that is delivered by the ECCS flow would (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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be true only after RCS pressure has dropped to very 2

low leyels late in the IOECCS scenario and long after 3

the non-escalation requirement has been violated.

4 High RCS pressures the critical or choked 5

water flow through the PSV will be much greater than 6

the ECCS flow that is delivered.

And I'm going to be 7

giving you another handout here.

This is an 8

illustration of this relationship for PSV.

9 The ECCS flow, I' 11 wait, this is what 10 I've been doing in retirement, been working on Excel 11 spreadsheets.

By the way, the ECCS flow that you see 12 here is conservative because it's high.

This is the 13 LOCA analysis people ref er to this as the max 14 safeguards case where all of the pumps are operating.

15 Where the ECCS flow is high and the relief 16 rate is low, the relief rate that you see here from 17 both the steam and the water, the water is two open 18 PORVs and the steam is also two open PORVs, stuck open 19 PORVs.

So this is an illustration of what would 20 happen if you're operating a

plant, Byron and 21 Braidwood would be one such plant, most plants do not 22 have safety grade PORVs.

23 If you operate such a plant and they pass 24 water they would stick open, both of them would stick 25 open not just one.

Here we have the red line is the (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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water relief rate.

And I say that is low, that is 2

conservative.

3 I'm using here saturated water

flow, 4

saturated critical flow.

If I were to use sub-cooled 5

flow you would multiply that by 150 to 200 percent.

6 So this gets the point across.

7 I' 11 also add at this point that other 8

statement in the FSAR, the one that says normal make 9

up systems, well ECCS is not a normal make up system.

10 The charging pumps when actuated by a safety injection 11 signal cannot be considered a normal make up system.

12 This discharging flow is not controlled by 13 a pressurizer level program or by let down flow rates.

14 It operates simply at maximum capacity and it does not 15 shutdown until the operator shuts it down.

That is 16 when they're actuated by an SI signal the function of 17 the charging pumps is to supply emergency core cooling 18 not to maintain a program pressurizing water level.

19 So they didn't even get that right.

So if 20 you look at this plot here I would suggest you read it 21 from right to left. Start with the high pressure. If 22 you look at the section between 2,000 and 2,500 PSI if 23 you go midway between those two there's 2, 250 PSI.

24 That's normal reactor operating pressure.

25 That's where you start times zero.

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that time you can see that the flow rate coming out of 2

two open PORVs, okay, this is steam flow.

It could be 3

due to anything.

4 Any time you reach that open pressure if 5

it happens at 2,250 PSI and I'm talking now about a 6

spurious opening of a PORV, if that happens at nominal 7

pressure, okay, then you might have, you can't keep up 8

with the flow rate through the PORVs with the Eees.

9 It's just not enough flow.

lb And if you start depressurizing and I'm 11 still looking at steam

now, if you start 12 depressurizing there's a crossover point at about 13 2,000 PSI.

At that point you might be able to keep up 14 with the flow with the Eees input.

15

However, by this time if you have an 16 inadvertent Eees actuation you're relieving water so 17 you're not down there with the steam release.

You're 18 up on the red line with the water release and you 19 can't, you're not, you can't possibly hope to make up 20 that flow with Eees.

21 And if you look at the dotted line, the 22 dotted line is just the charging parts.

So that would 23 represent the eves malfunction.

That's where a eves 24 malfunction is either due to an operator error or it's 25 due to something like a pressurizer water level sensor (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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that's failed, that fails low.

2 Okay, there are three pressurizer water 3

level sensors.

One of them controls the charging 4

flow.

If that one fails low it's going to send a 5

signal to the charging pumps to ramp up the flow and 6

try to make up and try to restore the water level and 7

you could have one or two charging pumps operating at 8

maximum capacity or close to maximum capacity due to 9

this failed sensor.

10 And this will continue until the operator 11 realizes what's happening and shuts that off.

So I 12 have a little discussion on the same plot here with 13 the situation that Exelon and Westinghouse both cited 14 just doesn't apply.

It doesn't apply to critical 15 flow.

16 It might apply very late in the transient 17 when the reactor cooling system is very low in 18 pres sure and it begins to 'look like a bucket.

You 19 know, you just throw water into a bucket and the water 20 that flows over the bucket is the same flow that goes 21 into the bucket.

22 By the time you get to those points you' re 23 long past the point where you violate the non-24 25 escalation requirement.

So what happens with the Byron and Braidwood plants and the Exelon licensing (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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rationale.

2 Basically Exelon is substituting safety 3

valves for PORVs to address the non-escalation 4

requirement.

To do this it has to show that the PSVs 5

will reseat after having relieved water.

6 Exelon claims it's safety valves are 7

qualified to relieve water based upon valve tests that 8

were conducted in 1988 by the Idaho National 9

Engineering Lab.

And I have, on the petition I might 10 have a section from this report.

11 Basically, I can read you the section but 12 basically I will tell you that what they did, shorten 13 it, what they did was Idaho National Engineering Lab 14 did not do the test and they say so in their report, 15 Section 4.2.3.

We didn't do this report because the 16 licensee, in that case it was Commonwealth Edison, 17 indicated that they had at least 20 minutes to shut 18 off the ECCS flow before the pressurizer was full and 19 that's plenty of time to do that.

20 So they didn't need to look at water 21 relief rates because the pressurizer wouldn't fill.

22 They didn't do the test.

But Exelon refers to that 23 anyway.

24 Also this report says that for a forward 25 plant which Byron plants are forward plants, both the (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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safety valves and PORVs will be challenged.

That's 2

not true, not unless you have it and you saw that in 3

the previous slide.

4 So when Exelon assumes the PSVs will 5

relieve water and then reseat it effectively imposes 6

two new design requirements on the PSVs.

Currently 7

the PSVs are designed to operate during Condition 4 8

accidents like f eedline breaks and beyond design basis 9

events like others where the reactor coolant system 10 pressure is the only issue you need to prevent all the 11 pressure, that's it.

12 That's all they do.

Once open the PSVs 13 will fulfill their RCS overpressure safety function.

14 It is not necessary to require that PSVs can relieve 15 water or even to reseat unless you want to use them 16 instead of the PORVs which Exelon wants to do.

17 So now the PSVs have new design 18 requirements.

Still when Exelon repurposed the PSVs 19 for use during AOOs and they could only use the PSVs 20 during AOOs if they used a time machine to get you 21 back to the AOO region because you' re already past it, 22 it became necessary to consider the possibility of a 23 PSV failing to close.

24 25 Now you get into GDC 21 requirements.

GDC 21 becomes a requirement for closure of PSVs.

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not a problem with opening the PSVs.

One failed open 2

PSV, okay, in other words one PSV that fails to close 3

if so would create a Condition 3 that would violate 4

the non-escalation requirement.

5 And this is a, this is simple.

The PSVs 6

are connected in parallel and they're not isolable.

7 This system can readily meet the GDC 21 single failure 8

requirement when the PORVs are required to open but 9

not, it cannot possibly meet it when the PORVs are 10 required to close.

11 So Exelon's plan to substitute PSVs for 12 PO RVs cannot meet the GDC 21 single failure 13 requirement.

I'm getting close to the end.

Don't 14 worry.

15 So you recall from what I said earlier 16 that the PORVs are designed to prevent unnecessary 17 challenges in the PSVs.

Exelon's compliance strategy 18 prevents the PORVs from opening and relies upon the 19 PSVs to open in lieu of the PORVs.

20 This creates a new accident.

For the 21 purpose of the petition I'm going to call this an 22 unnecessary challenge to the PSVs.

You see PSV.

This 23 is an AOO that pressurizes, this accident is an AOO 24 that pressurizes the RCS past the reactor trip safe 25 point to the PSV opening point and now you're up to (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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2,500 PSI which is the RCS design pressure.

2 And the frequency of this AOO is going to 3

be, again the sum of the frequencies of any AOO that 4

can pressurize a system for that long.

And then a PSV 5

as it works under these conditions as required by GDC 6

21.

7 So if you' re doing an analysis and even if 8

you qualified the PORVs for water relief and they're 9

already safety related, but if you qualify for water 10 relief you still need to consider that single failure 11 which GDC 21 requires.

So now you have three PORVs.

12 They're all open.

You need to close all 13 three PORVs.

If one doesn't close, I mean safety 14 valves, sorry.

So you've got three safety valves open 15 and they're qualified to relieve water.

They should 16 close, according to Exelon they should close because 17 they're qualified to relieve water.

18 You still need to consider one that 19 doesn't close to meet GDC 21 and you can't isolate it 20 anyway.

So safety significance.

So I'm looking at 10 21

CFR, Part 50.92 and that says issuance of an 22 amendment, Section A.

23 In determining whether an amendment to a 24 license will be issued to the applicant the Commission 25 will be guided by the considerations which govern the (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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issues of the initial license to the extent applicable 2

and appropriate. It also says the Commission may make 3

~ final determination that a proposed amendment to an 4

operating license involves no significant hazards 5

consideration if operation of it is still in 6

accordance the proposed amendment and there are three 7

questions here.

8 One, involve a significant increase in the 9

probability of consequences of a maximum previously 10 violated or two, create the possibility of a new or 11 different kind of accident from any accident 12 previously evaluated or three, involve a significant 13 reduction in safety margin.

14 And you've seen this in the petition.

I 15 would answer these questions in a way that differs 16 from Exelon.

All systems will not continue to be 17 operated in accordance with prior design requirements.

18 New design requirements have been imposed 19 upon the PSVs.

When operated during the IOECCS and 20 any other AOOs that pressurize the RCS for the PORV 21 opening set point, now the RCS will have to pressurize 22 to the PSV opening set point.

23 During each of these AOOs the PSVs will be 24 required to open and then reseat after having relieved 25 water.

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introduced, failure of a PSV to reseat.

2 If this occurs the result will be a small 3

hot leg LOCA and this LOCA is going to be more 4

frequent than the currently analyzed LOCA.

The 5

probability is going to be the sum of all those other 6

AOOs.

7 The consequences of the initiating AOOs 8

are also increased.

So that, since operation of the 9

PSVs at 2,500 PSI will always be required during an 10 incident in which the PORVs were currently open.

Not 11 when PSVs are open and not the PORVs.

12 The consequences of a stuck PSV if it 13 occurred would be greater than the consequences of a 14 stuck PORV.

And this is due to the fact that, first 15 of all you can't isolate a PSV and secondly a PSV is 16 twice the size of a PORV and there are three PSVs and 17 only two PORVs.

18 Operation of the PSVs during AOO is not in 19 design basis. Frequent pressurization of the RCS, and 20 this would occur during AOOs now you have frequent 21 pressure issue of the RCS to its design pressure, 22 2,500 PSI would also be outside the RCS design basis.

23 So in question two this is where you 24 create a new accident, an AOO that pressurizes the RCS 25 to the PSV opening set point.

And the frequency of (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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occurrence is the AOO frequency of occurrence.

2 I would note here that in the 1980s the 3

NRC was concerned what the safety implications of high 4

frequency of unnecessary automatic reactor trips that 5

will be incurred at operating plants.

I worked with 6

the Westinghouse Owners Group to reduce the incidents 7

of these trips.

8 And in the process I developed a system 9

that was patented. It's Patent Number 4, 832, 898.

You 10 can look it up.

You can just Google that number and 11 the patent will come up and it will show, it will show 12 that this patent was used to reduce the number of 13 unnecessary reactor trips particularly during start up 14 operations when feed and water control was manual and 15 it's also been referenced in four other patents.

16 This new accident is going to pose a 17 greater threat to the public health and safety of any 18 unnecessary automatic reactor trip because the 19 consequences of this event could be potentially 20 greater.

For example, the stuck open PORV would have 21 about twice the relief capacity, safety valve.

22 A stuck open safety valve would have about 23 twice the relief capacity as a PORV and it would not 24 be isolated and it could also exceed the number of 25 allowable pressurizations for the RCS.

When the PORVs (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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are applied the margin of safety is reduced.

2 This is the third question in the CFR Part 3

50.92.

You reduce the safety margin.

This is also in 4

the petition.

The safety margin right now is about 5

400 PSI.

That's 2,750 PSI minus the PORV opening set 6

point of 2,350 PSI.

7 However, when you pressurize to the level 8

of PSV opening now you've reduced the safety margin.

9 Now it's reduced to 250 PSI, 2, 750 PSI minus the 2, 500 10 PSI of the opening of set point for the safety valves.

11 So the margin of safety for this action 12 alone will reduce, is reduced by 38 percent.

And I'm 13 not going to, I'm going to skip forward a little bit 14 here.

So my conclusion is that Exelon's responses as 15 given in its most significant safety hazard statements 16 which conveniently are all negative.

17 I think just about every licensee submits 18 the most significant safety hazard statements and 19 lists all the answers negative.

These are not true.

20 And furthermore, Exelon' s responses may be false 21 statements of the kind that could violate the Federal 22 False Statements Statute, 18 USC Section 1001 Subpart 23 A.

24 According the court's conviction under 25 this act would require proof of five elements. First, (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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the defendant made a statement, okay.

Second, the 2

statement was false.

Third, the statement was 3

material.

4

Fourth, the statement was within the 5

jurisdiction of a government department or Agency.

6 And fifth, it's written and is made knowingly and 7

willingly.

So the first four elements I don't think 8

would be difficult to show.

9 But the fifth element that it requires 10 that the statement be made knowingly and willfully, 11 that could be hard.

But if the NRC improves its 12 communication with licensees, like for example 13 performing, taking the actions I asked for in the 14 petition making sure that licensees know what they're 15 writing into their FSARs and their LARs under oath and 16 affirmation.

17 If they know what's behind all of these 18 then a false statement could be construed as being 19 made knowingly and willfully.

So that's why it's 20 important to write a RIS and revise a RIS or write a 21 generic letter, revise the SRPs.

22 So in conclusion I would say that the 23 licensing basis of Exelon' s Byron and Braidwood plants 24 contains at least 12 errors.

Westinghouse shares in 25 some of these errors and the NRC fails to detect 11 of (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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them in at least three reviews.

2 So there's lots of blame to go around 3

here.

And I take my share of it becavse I was working 4

there when these things came through.

Exelon relies 5

upon the PSVs to open in lieu of the PORVs to relieve 6

water and then reseat.

7 Exelon claims that no PSVs will fail open 8

for water relief and therefore would not create a low 9

pressure clad of the pressurizer.

To reach this 10 conclusion it is necessary to repurpose the PSVs for 11 service during AOOs and to disregard GDC 21 single 12 failure requirement.

13 And to rely upon the PSVs Exelon assumes 14 the PORVs will not open.

A PORV that does not open 15 cannot fail open.

And so the non-escalation 16 requirement is satisfied by assumption.

17 Exelon' s conclusion begs the question. In 18 the process Exelon creates a new accident.

PSVs will 19 not open until after the RCS pressure during an AOO, 20 the pressure for which the AOO is defined, that is 21 after the high pressure reactor trip set point is 22 reached.

23 By the time the PSV is opened the AOO will 24 already have escalated to a Condition 3 event.

Exelon 25

. focuses upon qualifying the PSVs for water relief (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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duty.

And in order to substitute them for the PORVs 2

the qualifying valves that will not open before the 3

non-escalation criteria is violated makes no 4

difference whatsoever with respect to meeting the non-5 escalation criteria.

6 So in, my observation is that Exelon just 7

does not recognize the basic difference between PORVs 8

and PSVs and their respective functions in the plant.

9 And I'm going to use, I'm going to insult you with an 10 analogy here.

11 If the Byron and Braidwood plants were 12 automobiles then the PORVs would be seatbelts and the 13 PSVs would be airbags.

PORVs like seatbelts are used 14 often to protect the driver during abrupt stops, in 15 the occasional fender benders that could be engaged, 16 disengaged, disconnected even.

1 7 PSVs on the other hand are used once maybe 18 in a car's useful lifetime to protect the driver 19 during a head on collision.

Exelon's compliance 20 rationale does not and cannot demonstrate that its 21 Byron and Braidwood plant designs would prevent AOOs 22 from developing into more serious events.

23 Therefore, there is no assurance that 24 Condition 3 events will not occur and the frequency of 25 Condition 2 events in the Byron and Braidwood plants.

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Furthermore, Exelon cannot truthfully assert that 2

there are no significant hazards in these Byron and 3

Braidwood plants.

4 So I would request the NRC to take the 5

following actions.

Revoke Exelon's authorization to 6

operate the Byron and Braidwood plants at any uprated 7

power level.

Impose a license condition on current 8

operations requiring Exelon to provide an acceptable 9

demonstration of compliance with the aforementioned 10 design required.

11 And here I would refer you to the Seabrook 12 example of 2005 for a

precedent.

In that case 13 Seabrook was classified as operating under degraded 14 conditions and they were given until the next 15 refueling outage to correct it.

16 Require Exelon to file a 10 CFR Part 21 17 report and revise its no significant hazards 18 statement.

And I would also add that as far as I know 19 Byron and Braidwood plants are the only plants now 20 that are currently licensed to operate their safety 21 valves instead of their PORVs as a mitigation for the 22 ECCS actuation.

23 There is a li tiga ti on for the --

they 24 tried it and failed and used their PSV and even had 25 Westinghouse do an analysis for them and once they had (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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surprised them that they had the wrong discharge 2

coefficient and the water that was being relieved was 3

too cold to show operation of the safety valve water 4

relief they withdrew their application and instead 5

qualified their PORVs for water relief.

6 And the other uni ts are among the six 7

plants that are currently licensed to operate with 8

safety related water qualified PORVs not safety 9

valves.

And I think that is the proper way to go to 10 qualify PORVs.

11 In fact, I would go so far as to say 12 Westinghouse when they supplied these PORVs should 13 have supplied PORVs that were safety grade and go out 14 and qualify them.

But instead of making upgrades, 15 instead of operating the PORVs Westinghouse issued its 16 advisory in transferring that responsibility of doing 17 the work.

And qualifying PORVs is not the only 18 solution.

It's one solution.

19 And I would say that if this is going on 20 right now the licensing basis of the requirement of 21 the Byron and Braidwood plants.

It's also found in 22 the licensing basis of other plants. It's misleading, 23 maybe even false if you want to apply that willingly 24 and knowingly criteria it could be false.

25 (202) 234-4433 It's a continuing effect that needs to be NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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rectified because, I

believe too that look at 2

precedent of Byron and Braidwood qualifying their 3

safety valves for water relief I think there woulQ be 4

other licensees thinking to do the same.

And that is 5

the end.

6 MR.

WIEBE:

Okay, excellent.

Are you 7

tired?

We' 11 give you a chance to collect your 8

thoughts.

What was your time limit?

9 MR. MIRANDA:

I think it was one hour and 10 15 minutes.

11 MR. WIEBE:

We' re doing perfect.

So we' re 12 at the question phase.

And so our job is to seek to 13 understand.

And I might take advantage of you.

14 I' 11 ask some things that if you have 15 knowledge of it will be helpful.

But, you know, I 16 don't expect you to know all the answers to every 17 question that you get asked.

18 You

know, people are just trying to 19 understand.

So does anybody, you know, in the room in 20 the headquarters staff have any questions for Mr.

21 Miranda?

22 MR. OESTERLE:

I have one.

23 MR. WIEBE:

Okay.

24 MR. OESTERLE:

Eric Oesterle, Chief of the 25 Reactor Systems Branch in the NRR.

Well, Sam, I

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thought I

heard you say that you were with 2

Westinghouse for 45 years but you were not involved in 3

the development of the NSAL 93-13.

4 MR. MIRANDA:

That's right.

5 MR.

OESTERLE:

You weren't involved in 6

developing.

Were you involved in approving it?

7 MR. MIRANDA:

I was not involved in any 8

way.

9 MR. OESTERLE:

Any way, okay.

Thank you.

10 MR. WIEBE:

Okay~ I'll go.

Help me with 11 this one.

You know, the remedy you asked for is, you 12

know, revoke the authorizations to operate at the 13 uprated power.

14 So it's probably a two part question.

Why 15 do you feel comfortable with Byron and Braidwood at 16 their normal power?

And does a power uprate really 17 have anything to do with an inadvertent operation of 18 ECCS?

19 MR. MIRANDA:

Okay, first of all I'm not 20 comfortable with them operating at the current power 21 level and otherwise would have said so.

However, I'm 22 looking at the precedent for Seabrook.

23 Seabrook is looking for a scope power 24 uprating and they had similar problems.

They were not 25 trying to qualify their safety valves.

But they came (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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in with a license amendment request that had an 2

analysis that indicated that their pressurizer would 3

fill in something like 9.9 minutes and the standard 4

that the NRC was using was 10 minutes.

5 So if you could show the pressurizer 6

filled in ten minutes or greater than it was 7

acceptable.

So here they were, no, it was 10.1 8

minutes.

They were just over the line.

9 Their analysis looked like it was fine 10 tuned to meet that criteria.

And I just, I didn't buy 11 it.

So I asked them some questions and said how are 12 you going to meet this non-escalation criterion and in 13 the end they were given that scope power uprating 14 provided they could meet the accepted criteria in the 15 non-escalation criterion within the next power, before 16 the next power outage.

17 So I'm not really happy with them 18 operating at their current power level.

But I'm not 19 going to sit here and ask you to shut them down.

But 20 they could do at least what Seabrook did and you come 21 up with an acceptable, within a reasonable time, an 22 acceptable way to show compliance with the design 23 requirement that's been in place now since 1973.

24 25 MS. KIRKWOOD:

I have a question. Are the consequences worse at the higher power level?

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MR.

MIRANDA:

Technically, you

know, 2

that's, for the inadvertent Eees actuation technically 3

it should not be affected very much if at all on the 4

higher power level.

However, if you look at the eves 5

malfunction they attribute and when I say that it's 6

because the difference between these two remarks is 7

Eees actuation involves an immediate reactor trip.

8 As soon as you get that SI signal the 9

reactor trips.

In a eves malfunction the reactor 10 doesn't trip.

You rely on something else to trip the 11 reactor.

Meanwhile you' re generating power at the 12 higher power level and after the reactor trip you have 13 the decay heat in the system that is consistent with 14 the higher power level.

15 With a higher power level, in my opinion, 16 is going to have an effect.

But considering, depends 17 on how much higher the power level is.

Exelon, for 18 example, asked for two upratings.

One is a stretch 19 uprating is one something less than five or six 20 percent and the other one was a measurement uncertain 21 with a certainty of recovery.

22 And that was, you

know, probably one 23 percent or something like that.

It's a small affect 24 but it's there and the affect is due to the post-trip 25 decay heat generation.

The post-trip decay heat (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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generation is going to cause the

reactor, the 2

pressurizer to fill faster.

3 MS. KIRKWOOD:

I also heard of the plan 4

too that they failed to meet the non-escalation 5

requirements.

What are you referring to?

6 MR. MIRANDA:

That is the, there are three 7

requirements that all AOOs must meet.

These are the 8

Condition 3 events which are defined with events that 9

could occur any time during a

reactor year of 10 operation.

11 So you could have, in fact when I was 12 working on the trip reduction program at the 13 Westinghouse Owners Group there were plants back in 14 the 80s that were tripping, they were tripping maybe 15 six or seven times a year, unnecessary reactor trips.

16 And an unnecessary reactor trip, by the 17 way, is defined as an AOO, Condition 2 event.

So the, 18 for Condition 2 events basically the requirement is, 19 there are three requirements, lower the pressure, no 20 fuel cladding damage and not allowing the event to 21 become worse, to become a Condition 3 alert.

That's 22 the non-escalation.

23 MS. KIRKWOOD:

But where is that, where 24 are those requirements recorded?

25 (202) 234-4433 MR.

MIRANDA:

They are coming from a NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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standard that was written in 1973 by the American 2

Nuclear Society and which was adopted by the industry 3

and repeated in the FSARs.

It says up front in 4

Chapter 15 they're going to give you analyses of these 5

AOOs and we' re going to show that they meet these 6

three requirements.

7 It was in the standard.

That's where it 8

originated. But then a standard is not a requirement.

9 But when they put it into the FSAR and they tell you 10 that they're going to meet those requirements and the 11 NRC issues an operatirig license based on them meeting 12 those requirements I would say that's a requirement.

13 MS. KIRKWOOD:

So it's in the FSAR?

14 MR. MIRANDA:

Yes.

15 MS.

KIRKWOOD:

In the final agreement 16 FSAR?

17 MR. MIRANDA:

Yes, it is.

18 MR. WIEBE:

Can I do a follow up on that 19 one?

So do you know specifically where they endorse 20 the AMS standard as opposed to, Chapter 15 they have 21 the three criterion written out.

22 I can just go and look at them.

Do you 23 know where they endorse that in the standard, in the 24 FSAR?

25 MR.

MIRANDA:

You mean Exelon or (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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Commonwealth Edison?

I don't have their documents.

2 I wouldn't know where they put that.

I don't know 3

even if they don't.

4 But I do know that a lot of licensees like 5

Exelon reference that standard in their FSAR.

So they 6

will put those three requirements in Chapter 15 and in 7

the references they'll have the AMS standard.

8 MR.

WIEBE:

So for the record he's 9

referenced three in the petition, that's the edition 10 he's referring to.

11 MR.

CASE:

Is there any difference 12 between, you reference two editions, is there any dare 13 there?

14 MR. MIRANDA:

Referenced what?

15 MR. CASE:

You referenced two different 16 editions of the ANS standard.

Is there any --

17 MR. MIRANDA:

Yes.

18 MR. CASE:

The '73 and then the '83.

19 MR.

MIRANDA:

They're used 20 interchangeably.

But they're not.

They're different 21 standards. The 197 3 standard established, defines the 22 Condi ti on 2, 3 and 4 events and also defines the 23 acceptance criteria and also gives examples of what 24 these are.

25 (202) 234-4433 And if you look at that ANS standard in NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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1973 and if you look at Reg Guide 1. 70 which was 2

issues, I think first issued in 1968 you will see a 3

lot of similarities.

51.1 which came ten years later 4

has more conditions.

5 Now there are five conditions.

Nobody 6

really adopted that, okay.

It's really complicated.

7 And I think what my guess is that the 1983 standard 8

was issued because the ANS has a procedure that says 9

if a standard is issued and it's not revised within 10 ten years then it's withdrawn.

11 So here you have the ANS '73 standard and 12 developed '83.

They decided to embellish upon it. As 13 far as the NRC is concerned I don't think it matters.

14 Once the licensee commits to meeting these three 15 standards it doesn't matter where they come from.

16 It's like playing pool.

You call your 17 shot and you make it.

These are the requirements 18 we're going to meet.

Here are the analyses.

We've 19 met them.

The NRC, as you know, looks at that and 20 says, yes, you've got your license.

21 PARTICIPANT:

One more question. What do 22 you see as the difference between --

23 (Telephonic interference.)

.2 4 25 mention --

(202) 234-4433 MR. MIRANDA:

I was very careful not to NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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(Telephonic interference.)

2 PARTICIPANT:

I noticed and that's fine.

3 MR.

MIRANDA:

Then there's other 4

implications. The difference is that although backfit 5

applies to the Byron and Braidwood plants and this is 6

much more encompassing.

7 There are lots of, this covers all PWRs.

8 This third criterion, the non-escalation criterion is 9

the criterion that's hardest to meet.

And there might 10 be 40 or 50 in operation right now.

11 They have different ways of meeting the 12 standard.

And I will say that of all of the PWRs 13 maybe a

quarter of them adequately meet this 14 requirement.

The rest have various technical errors 15 and in the case of Byron and Braidwood its thermal and 16 technical is logical.

17 You know, they somehow are begging the 18 question because of my rationale.

The backfit, the 19 implication of the backfit is I kind of touched on it 20 earlier, it sets a precedent.

21 The Byron and Braidwood plants if Exelon 22 succeeded in licensing four plants in Illinois based 23 on begging the question.

So why can't other licensees 24 do the same?

25 (202) 234-4433 And I would say probably if I had to point NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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to an example that I would like to see I would like to 2

see licensees follow the example of Millstone, 3

Millstone Unit 3.

Millstone Unit 3 qualified their 4

PORVs as safety grade equipment, qualified them for 5

water relief and also had Westinghouse modify their 6

reactor protection system such that the units worked 7

in ECCS actuation event.

8 It's no longer a possibility.

They have, 9

what they did, Dominion went to Westinghouse and said 10 we would like you to change our reactor protection 11 system so now if we have a safety injection signal 12 from whatever the source, Millstone had an event like 13 this occur in 2005 in which case it had a tin whiskers 14 growing in one of their circuits in their protection 15 system.

16 You see this here, it's completed when 17 necessary for a reactor, a safety injection system.

18 So a safety injection system sparks and fills the 19 pressurizer.

The relief valves open, relieve water.

20 In that case they had reseated.

21 But after they reseated them.

So they 22 went to Westinghouse and said we want to change the 23 reactor protection system so that now a

safety 24 injection signal alone is not sufficient to start the 25 safety injection system.

It takes, for Millstone it (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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takes a safety injection signal plus a low pressurizer 2

pressure signal, both.

3 And the low pressurizer pressure signal 4

opens the valves that admit ECCS water into the 5

reactor coolant system.

So you have a coincidence 6

there.

You need both.

7 MR.

CASE:

Okay.

We'll get to our 8

participants on the phone and then we'll come back 9

here and if there's more questions we can do those.

10 So how about some of the headquarters staff that are 11 on the phone?

Do you have any questions for Sam?

12 Okay, hearing none you can always chime in 13 later.

Our marvelous regional staff, does our 14 regional staff have any questions for Sam?

My 15 questioning inspectors have no questions.

Okay, was 16 the licensee on the line?

17 PARTICIPANT:

How about the public?

18 MR.

CASE:

Yes, we're getting there.

19 Thank you.

Does the licensee have any questions for 20 21 22 23 24 25 Mr. Miranda?

Okay, good.

Now we can move to the public.

How about members of the public, any questions?

MR. LEWIS:

My name is Marvin Lewis.

May I speak now?

(202) 234-4433 MR. CASE:

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62 1

can only ask questions about the process.

So fire 2

away.

3 MR. LEWIS:

The question I wanted to ask 4

was Sam's telephone number or email so I could get a 5

hold of him.

If I can't ask that question I'll find 6

somebody I can ask.

7 MR. CASE:

I think Joel is going to get 8

your name and number and email.

And of course 9

although the records will be public and Sam can shoot 10 it out to you.

11 MR.

LEWIS:

Okay.

Then may I ask. a 12 question on, anything else that would be technical.

13 Okay, so I'll go with APA and worry about technical 14 stuff when I get Sam's number, Mr. Miranda's number.

15 I forget his exact name, I'm sorry.

16 MR. WIEBE:

Marvin, this is Joel.

Sam's 17 number is on the petition, the first page of the 18 petition at the bottom of the page.

19 MR. CASE:

Sam, what were you thinking?

20 MR. LEWIS:

Okay.

I haven't been able to 21 find it. But that's all right. What's your number so 22 I can get a hold of you?

23 MR.

WIEBE:

Yes, we can.

We walked 24 through that with you, I walked through that with you 25 earlier that Adam's number.

That was the petition and (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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it's on the bottom of the page.

2 MR. LEWIS:

Very good, okay.

So I'll look 3

it up.

Thank you.

4 MR. CASE:

If you can't find it we'll do 5

it again.

6 MR. LEWIS:

Excellent.

7 (Off microphone comments) 8 MR. CASE:

Okay, Marvin, do you have any 9

more comments?

10 MR.

LEWIS:

No, it sounds exactly like 11 Three Mile Island where I won a contention as a pro se 12 intervener.

The only time a pro se intervener ever 13 won ~ contention with, in an operating nuclear power 14 plant, meaning Three Mile Island Number 1 restart, not 15 Number 2.

16 Number 2 ain't restarted, thank heavens.

1 7 But I did get the vent on the Number 1 before it 18 restarted. And Number 1 hasn't followed in Number 2' s 19 tracks.

I wonder how much was because of the vent I 20 got on there.

21 Well, thank you.

Thank you for allowing 22 me to make my comments.

23 24 25 MR. CASE:

Okay, now do we --

MR.

BORROMEO:

Sam, can you expand on a little bit why, this is Josh Borromeo, why it's non (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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conservative to assume the PSV opens as opposed to the 2

PORV when trying to meet the non-escalation criteria?

3 I would think PSV is a bigger hole, right.

4 Why is the, what assumptions are you 5

making there?

6 MR. MIRANDA:

Well when you do an analysis 7

of an IOECCS malfunction you' re testing whether, first 8

of all you want to know how quickly the pressurizer 9

will fill.

If it takes a long time like in 1988 10 Commonwealth Edison was telling people that it took 20 11 minutes to fill the pressurizer.

12 It takes a long time then there is an 13 issue because 20 minutes is plenty of time to prevent 14 that.

But when the pressurizer fills the first valve 15 that's going to open is the PORV.

16 MR.

BORROMEO:

So it's the timing of it 17 versus the size of the hole?

18 MR. MIRANDA:

It's, you have PORVs and you 19 have PS Vs neither of which is qualified to relieve 20 water.

And you're testing in the analysis if you're 21 going to open a valve relieve water and have it stick 22 open.

23 So the first valve that opens and relieves 24 25 water is the PORV.

Now Westinghouse in their NSAL made a common mistake which is that a PORV, since it's (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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not a safety-related component, it's control grade, 2

it's just assumed it doesn't operate.

3 That's not the whole story.

The whole 4

story is you assume it operates if it makes the 5

accident worse and you assume it doesn't operate, 6

normally you assume it doesn't operate.

7 But if it's operational and makes the 8

accident worse then you assume it does operate.

In 9

this case, if you open the PORV and have it relieve 10 water now you raise the possibility that it could 11 stick open.

12 First, it opens, actually both PORVs will 13 open.

14 MR. BORROMEO:

Both PO RVs will open first?

15 MR.

MIRANDA:

And they will both stick 16 open because neither of them is qualified to relieve 17 water.

18 MR. BORROMEO:

Sam, can I do one more? It 19 relates to your question.

This issue has a long 20 history.

So when I'm reading the petition what's the 21 new information that I should focus on?

22 There's all this, you know, information.

23 Is this all in the record too?

24 25 MR. MIRANDA:

I have a book at home about this.

The new information that I would, I would say (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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the new information is that the old information is 2

licensees have a lot of, historically have had a lot 3

of trouble meeting this criterion, requirement.

4 The new information is that one licensee, 5

Exelon does succeed in its license, has succeeded in 6

convincing the NRC that qualifying the safety valves 7

for water relief is sufficient to demonstrate 8

compliance with the non-escalation requirement.

9 And what I'm saying here is it doesn't 10 matter whether the safety valves are qualified for 11 water relief or not.

It does nothing whatsoever to 12 advance some kind of demonstration of compliance for 13 the non-escalation design requirement.

14 Safety valves are not supposed to be 15 operating.

They're not supposed to open at all.

You 16 can't substitute safety valves for PO RVs.

They're 17 different components.

They' re there for different 18 reasons, neither of which is qualified for water 19 relief.

20 But about a half a dozen plants have 21 succeeded in qualifying the PORVs for water relief.

22 That's been accepted by the NRC and in fact, that 23 makes sense.

That makes sense.

24 25 But qualifying safety valves for water relief doesn't makes sense because by the time they (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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open 2,500 PSI you have already gone beyond the AOO.

2 You've already violated the criterion.

3 MR.

BORROMEO:

Did we talk about that 4

before, you know?

In this history did we talk about 5

6 MR. MIRANDA:

I don't think we have.

I 7

don't think we have.

I think engineers like, have had 8

to learn to admit *and I'm guilty of this too that it's 9

an interesting problem.

10 You have a huge safety valve, 3.6 square 11 inch flow area, the largest valve and then it's not 12 qualified for water relief. It would be nice to have 13 it qualified for water relief because you know it's 14 going to open during a feedline break and relieve 15 water.

16 So there are valve tests out there to try 17 to get this qualified for water relief.

I don't think 18 any of them have been successful.

So this is one, 19 this is a problem you could work out for years and try 20 to get this solved.

21 But in the end, for this verification or 22 demonstrating that you meet the non-escalation 23 requirement it makes no difference whatsoever because 24 that safety valve should not even be open.

25 MR.

BORROMEO:

I have some follow up (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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questions.

So getting to your actions requested in 2

the petition you asked that the NRC revoke the 3

licensees authorization to operate Byron and Braidwood 4

at any uprated power level and to impose a license 5

condition similar to Seabrook I think it was.

6 MR. MIRANDA:

Yes.

7 MR. BORROMEO:

Which was a full reanalysis 8

of the IOECCS events and using NRC approved 9

methodology to qualify the PORVs for water relief.

10 Would you be okay with the second remedy for Byron and 11 Braidwood and still allow them to operate at the 12 uprated power level?

13 So I'll ask it another way.

If Bryon and 14 Braidwood uprated their PORVs for water relief would 15 it be safe for Byron and Braidwood to operate at the 16 uprated power levels?

17 MR. MIRANDA:

I believe that if Byron and 18 Braidwood qualifies their PORVs for water relief and 19 that involves more than just water relief, it involves 20 qualifying them as safety grade components, if they do 21 that it puts them in the same category as Millstone, 22 Diablo Canyons, Salem and I think at least a half a 23 dozen plants have done that and that's been acceptable 24 and I would be okay with that.

25 (202) 234-4433 And I would also add that the Seabrook NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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license con di ti on, if you want more information on 2

that, the project manager at the time was Gerald.

In 3

fact, it was his idea.

4 MR. BORROMEO:

My second question is this.

5 You said you were involved in developing RIS 2005-39 6

that was on the topic.

In your experience in doing 7

that how consistent were licensees and the NRC in 8

approving for identifying IOECCS events as Condition 9

2?

10 Were they all Condition 2 AOOs or sort of 11 some licensees treat them as Condition 3 accidents?

12 Was there consistency or was there variation?

13 MR. MIRANDA:

No, they were all considered 14 Condition 2 events.

And I think if you look at Reg 15 Guide 1.70, Table 15 you should find it listed as one 16 of the events that's Condition 2.

17

Yes, Table 15-1, Section 5 increasing 18 reactor coolant inventory, Section 5.1 and 5.2, known 19 as ECCS actuation.

There are eves malfunctions.

20 There was never any question that would be 21 a Condition 2 event from that standpoint, from the 22 regulatory standpoint.

And I would say also from the 23 operating experience standpoint.

24 Probably the most frequent event after an 25 unnecessary reactor trip would be the inadvertent ECCS (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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actuation.

And sometimes they come together.

2 PARTICIPANT:

I'm Tim Drzewiecki.

I'm up 3

in the Off ice of, Reactor Systems Branch.

So my 4

question is about the opening of an SI.

What's the 5

basis saying that the event would be non-escalation?

6 Does that come other than the standard?

7 MR. MIRANDA:

Well that was my Figure 1.

8 I think, did you get Figure 1?

Yes, I think you came 9

in late.

10 PARTICIPANT:

The 73 standard that you're 11 pointing too, okay.

12 MR. MIRANDA:

If you look at that figure, 13 yes, it's that 73 standard and you see that the safety 14 valve doesn't open until after the AOO has become 15 trips.

2,500 PSI and then to show the trip at 2,400.

16 But if the reactor trips at 2,400 PSI and 17 the pressure is still increasing it means that event 18 is not accommodated by a reactor shutdown.

There's 19 something else going on.

20 MR.

BORROMEO:

So,

Sam, this is Josh 21 Borromeo again.

So the, I'm just trying to square 22 this in my head.

So the, so IOECCS is initiated by a 23 reactor trip?

24 25 MR. MIRANDA:

Sometimes or sometimes it happens, it's called a trip with complications.

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MR.

BORROMEO:

So trips can still have 2

IOECCS, right and then --

3 MR. MIRANDA:

Yes.

4 MR.

BORROMEO:

So then if there's an 5

inadvertent complicated trip would it automatically 6

become a Condition 3 event?

7 MR. MIRANDA:

No, I don't believe so.

In 8

fact, this is what happened at Salem in 1994.

They 9

had a reactor trip and along with it came an ECCS 10 actuation which filled the pressurizer and opened the 11 PORVs.

12 What happened there was you had a

13 Condition 2 event.

You have the reactor trip and a 14 second Condition 2 event, ECCS actuation.

And that 15 second Condition 2 event may have occurred because of 16 the reactor trip or independently.

17 It doesn't matter.

But it's still a 18 condition for the event.

19 MR.

BORROMEO:

Yes, I'm just trying to 20 square it with the figure right.

So like if it 21 doesn't, if the reactor doesn't trip it's a Condition 22

2.

But if it does it's a Condition 3 based on Figure 23 1, right?

24 That's what's, so if you fill your 25 pressurizer it goes above 2,400 PSI the reactor trips (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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and then you're saying it becomes a Condition 3 then?

2 MR.

MIRANDA:

It becomes, the standard 3

says that an AOO shall be accommodated by a reactor 4

shutdown.

A shutdown is at 2,400 PSI, I'm not talking 5

about an inadvertent ECCS actuation.

6 I'm talking about any AOO, anything that 7

pressurizes the reactor coolant system.

If it 8

pressurizes the 2,400 PSI and the pressure continues 9

to increase it means that reactor shutdown could not 10 accommodate the AOO.

Therefore, it's not an AOO.

11 It's something else.

12 MR. BORROMEO:

Okay.

I understand.

So it 13 should have shut it down prior to 2,400, right?

14 MR. MIRANDA:

Yes, like for example loss 15 of feedwater that's a very common AOO.

So with the 16 loss of feedwater they cause the reactor system to 17 pressurize.

18 As an insurge into the pressurizer the 19 pressurizer water level goes up and some times during 20 this period the reactor trips.

And after the reactor 21 trip it channels into the feedwater the decay heat at 22 first is very high, auxiliary feedwater should assume 23 a single player is not going to take care of all of 24 the decay heat.

25 (202) 234-4433 So the pressurizer water level will NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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continue to increase with time.

But eventually the 2

heat generation will be removed by the feedwater 3

system and then you'll reach a peak level with the 4

pressurizer of water level then it will start to go 5

down.

6 And from then on now and forever it's just 7

recovering the plant.

8 MR. WIEBE:

Okay.

Any other questions?

9 Okay, Sam, thank you for taking the time to provide 10 the NRC staff with clarifying information on the 11 petition you submitted.

12 (Whereupon, the above-entitled matter went 13 off the record at 2:47 p.m.)

14 15 16 17 18 19 20 21 22 23 24 25 (202) 234-4433 NEAL R. GROSS COURT REPORTERS AND TRANSCRIBERS 1323 RHODE ISLAND AVE., N.W.

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CERTIFICATE This is to certify that the attached proceedings before the United States Nuclear Regulatory Commission Proceeding: 10 CFR 2.206 Petition Review Board Conference Call RE Braidwood/Byron Docket Number: N/a Location: Teleconference were held as herein appears, and that this is the original transcript thereof for the file of the United States Nuclear Regulatory Commission taken and thereafter reduced to typewriting under my direction and that said transcript is a true and accurate record of the proceedings.

Official Reporter Neal R. Gross & Co., Inc.

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WASHINGTON, D.C. 20005-3701 www.nealrgross.com