PRM-50-93, no sweat at NRC's OGC

So, here is the latest. Click to enlarge, and use your back arrow to return. Mote later regarding this.

CHAIRMAN ABDEL-KHALIK: "You have five minutes to make your remarks."

Forward: On January 14, 2011, I addressed the ACRS in matters related to the Rod Bundle Heat Transfer program at Pennsylvania State University and also the User Need Request that is on file (ML100770218). Although ACRS allows NRC staff to use all of the time they want in generally poorly organized presentations, I was restricted to five minutes. Following is copied from the transcript of the meeting. I participated via a telephone connection. My two page handout is also here.

CHAIRMAN ABDEL-KHALIK: Mr. Leyse?
MR. LEYSE: Yes, can you hear me?
CHAIRMAN ABDEL-KHALIK: Yes, we can. You have five minutes to make your remarks.
MR. LEYSE: Well, do you want me to start now?
CHAIRMAN ABDEL-KHALIK: Yes.
MR. LEYSE: Okay. It'll take me 10 seconds to walk to my notes, and we'll be going.
I’m Bob Leyse. I have 5 minutes. Starting with slide 1:

On December 2, 2010, I taught Full ACRS that 2200 degrees Fahrenheit is too high as a PCT to insure that thermal runaway will not occur in a LOCA.

Today I’ll focus on two items:

RBHT at Penn State and the User Need Request, Leeds to Sheron, April 26, 2010, which is a user need request for a technical analysis of PRM-50-93. November 17, 2009.

Starting with RBHT:

RBHT has apparently explored the relatively low temperature regions of LOCAs utilizing its 49 rod full length assembly with Inconel clad heaters. Very likely millions of dollars have been spent over the 13 year activity. The most recent pubic discussion of RBHT was at the ACRS Thermal Hydraulic Phenomena Subcommittee, Monday, October 18, 2010. Of course, the general uselessness of RBHT is due to its lack of data with zirconium alloy cladding in the region greater than 1800 degrees Fahrenheit that is documented in plant licenses. NRC has avoided exploring this region with multirod assemblies having zirconium alloy cladding.

NRC (outrageously) has always promoted RBHT at Penn State
as highly applicable to TRACE and licensing. However, the documents are generally not available to anyone outside of NRC and its contractors.

ACRS Consultant Wallis may have had such access because at the cited meeting of the Thermal Hydraulic Phenomena Subcommittee he observes: “Are we going to hear about this later? Because the only thing I have seen from the Penn State work was some very sort of crude results,
but they measured all kinds of stuff.”
Next, I’ll move to the user need request, I’ll cite a tie-in to Penn State’s RBHT.

In the User Need Request, Leeds to Sheron, April 26, 2010, Leeds refers to the Technical Safety Analysis dated April 29, 2004, of my PRM-50-76, docketed May 8, 2002, as an “… outstanding technical analysis … .” However, the facts reveal that NRC’s Technical Safety Analysis of PRM-50-76 is most certainly not an outstanding technical analysis. Referring to RBHT, the Technical Safety Analysis of April 29, 2004 reports, “Current programs at Pennsylvania State University … are far more cost effective.”

So, in 2004, NRC staff was praising RBHT, but more than 6 years later, Expert Consultant Wallis reported, “…Penn State work was some very sort of crude results.”

Now, since RBHT has only used inconel clad bundles, it is absurd that Leeds lauds the 2004 Technical Safety Analysis of PRAM-50-76 as an “… outstanding technical analysis … .”
Of course there is much more documentation of the defects in RBHT and the User need letter than I am covering in 5 minutes.

Slide 2 has blue and black type. The blue type is what the Thermal Hydraulic Subcommittee was told on October 18, 2010, as its list reports that are dated 2008. The black type reveals that none of the reports have been released by NRC and three of the reports have no assigned date of release.

I still have over one minute. It’s not on either slide, but in the referenced meeting of the thermal hydraulics subcommittee there was a lot of discussion of the impact of various grid features, such as mixing vanes, on test results. However, if zircaloy grids had been used for comparison with inconel, and if the tests were conducted at realistic temperatures depicted in actual plant licenses, the impact on test results would have been far greater than the relatively minor impact of mixing vanes.Finally, I should not have been restricted to five minutes, more later on that. That’s it.

CHAIRMAN ABDEL-KHALIK: Thank you, Mr. Leyse. Are there any questions for Mr. Leyse? Well, hearing none, thank you. Are there any additional questions to either the staff, or to AREVA considering the fact that this is an open session? Hearing none, we will recess at this time. Our schedule calls for us to go to a lunch recess. We will reconvene at 1:15, and at that point we will be off the record.

Following are the two slides that are in the transcript. Click on the slide to enlarge and use your back arrow to return here.

Some history -- Microscale Heat Transfer to Subcooeld Water, 200-6000 PSIA

Microscale Heat Transfer to Subcooled Water
200-6,000psia, 0-3,500W/cm2
ROBERT H. LEYSE Article first published online: 24 JAN 2006

DOI: 10.1111/j.1749-6632.2002.tb05912.x
Issue
Annals of the New York Academy of Sciences
Volume 974, MICROGRAVITY TRANSPORT PROCESSES IN FLUID, THERMAL, BIOLOGICAL, AND MATERIALS SCIENCES pages 261–273, October 2002

LEYSE, R. H. (2002), Microscale Heat Transfer to Subcooled Water. Annals of the New York Academy of Sciences, 974: 261–273. doi: 10.1111/j.1749-6632.2002.tb05912.x

Author Information
Inz, Inc., Sun Valley, Idaho, USA
*Correspondence: ROBERT H. LEYSE,

*Correspondence: Address for correspondence: Robert H. Leyse, Inz, Inc., P. O. Box 2850, Sun Valley, ID 83353, USA. Voice/fax: 208-622-7740; Bobleyse@aol.com.

Publication History
Issue published online: 24 JAN 2006
Article first published online: 24 JAN 2006

Keywords:microscale heat transfer;ultra-high heat flux;supercritical heat transfer;phase change heat transfer;subcritical boiling heat transfer;intense turbulence

Abstract: Exciting heat transfer phenomena have been discovered with a micron-sized heat transfer element operating in subcooled (20°C) degassed, demineralized water over a wide pressure range (200-6,000psia) at heat fluxes up to 3,500W/cm2. The platinum heat transfer element (diameter 7.5 microns, length 1.14mm) is installed within a one-cm3 stainless steel chamber. Sealed electrical terminals penetrate the chamber to effect direct current heating of the platinum element. Pressure is applied pneumatically. The adiabatic heating rate of the element is 6°C per microsecond at 3,700W/cm2; response is essentially instantaneous for the procedure described herein. The direct current voltage and current are measured from which the power and the resistance (temperature) are determined. The following procedure applies: (1) Pressurize the water-filled stainless steel chamber to 6,000psia. (2) Apply power at 3,000W/cm2. (3) Maintain constant heat flux as pressure is smoothly reduced from 6,000psia to 200psia over a period of 20 seconds. Record voltage, amperage, and pressure at 0.1 second intervals. Heat transfer phenomena thus discovered: (1) Element starting temperature of 370°C at 6,000psia smoothly increased to 380° as pressure was reduced to 3,970psia. (2) At 3,970psia the temperature abruptly stepped upward to 590°C. (3) Temperature smoothly increased to 730°C as pressure was reduced to 3,230psia. (4) In the vicinity of the critical pressure, the temperature turned around and began smoothly decreasing. (5) At 2,350psia, the temperature stepped down from 520 to 350°C. (6) Temperature smoothly decreased to 230°C at 190psia and power was then turned off. Bulk water temperature increased less than 4°C. Controlled gravity (KC-135) tests are planned.

e-mail to NRC OGC Stephen Burns

I sent this to Burns on February 4, 2011. I’ll report the response.

Bajorek Reference to OGC: ACRS Subcommittee, October 18, 2010

Sir:

Please fill me in on the details of communications between Bajorek and perhaps others and yourself and/or your staff regarding the timing of the NRC’s evaluation of PRM-50-93. I was quite surprised to learn at the October 18, 2010, meeting of the Advisory Committee on Reactor Safeguards Thermal Hydraulic Phenomena Subcommittee that you advised that the evaluation of PRM-50-93 could be delayed. According to Bajorek, you advised that the date of August 2010, that is specified in the User Need Request, Leeds to Sheron, April 26, 2010, ML 100770117, could be delayed by about as long as the NRC decided to delay that evaluation. Bajorek reported, “But because they amended their own petition, and submitted
another petition, OGC decided to lump it together and
that window of time has moved out.”

Following is from the transcript of that meeting:

Page 468 follows:
16 CONSULTANT KRESS: I found it very unusual
17 that public comments are made to the subcommittee.
18 Those usually go to the full committee. I don't know
19 what your obligation is with respect to those.
20 CHAIR BANERJEE: I think to report it to
21 the full committee and ask if --
22 CONSULTANT KRESS: Just report it to the
23 full committee.
24 CHAIR BANERJEE: ask if they wish it to be
25 made to the full committee. I don't think that we can
act on it.
2 CONSULTANT KRESS: No. That was my point.
3 It has to be acted by the full committee.
4 CONSULTANT WALLIS: But if you want a
5 comment, it looked as if there could be a significant
6 point here, I mean it's something that is not trivial
7 to look at and see is there a question here and what's
8 the evidence for --
9 CHAIR BANERJEE: Has the comments been made
10 to the staff or is it just to the subcommittee?
11 MR. BAJOREK: This is Steve Bajorek.
12 Actually there are two petitions in play right now.
13 The petition they talked about brings up the point
14 that they Baker-Just is possibly not conservative. He
15 has the same comment on Cathcart-Pawel. Asks to look
16 at some of these other test data that he claims we
17 have not looked at before.
18 He also submitted --
19 CHAIR BANERJEE: Particularly bundle data.
20 MR. BAJOREK: Bundle, yes. The staff has
21 put together a small group to start to evaluate these
22 concerns. We started to take a look at it and another
23 petition came in, this one on the behalf of
24 Connecticut or Yankee, it's a plant that's been up for
25 relicensing. There are --
CONSULTANT WALLIS: Vermont Yankee?
2 MR. BAJOREK: Vermont Yankee, that's right.
3 Vermont Yankee is being relicensed. They have also put
4 in a petition on their behalf where they cite many of
5 the same concerns. Because these petitions are over
6 lapping, the staff decided they were not going to look
7 at them individually, they were going to put them
8 together. We went through our OGC. They said that was
9 an appropriate thing to do and now the window of time
10 for evaluating those petitions and those concerns has
11 been reopened and I think we have another -- I think
12 we have a year to go through and reevaluate
13 everything. So there's a group that is looking at
14 that.
15 CHAIR BANERJEE: So I think we can report
16 that to the full committee.
17 CONSULTANT WALLIS: But just report that.
18 That's all we have to do.
19 MEMBER ABDEL-KHALIK: And I think from the
20 committee's perspective, we await the staff's
21 evaluation and we will review the staff's evaluation.
22 MR. BAJOREK: He did make the point that
23 while there was a user need letter, point out and the
24 research was supposed to have responded by I think the
25 end of August. That was the original schedule. But
because they amended their own petition, and submitted
2 another petition, OGC decided to lump it together and
3 that window of time has moved out.
4 CHAIR BANERJEE: Okay. Well with that, I
5 think I'd like to thank you all and adjourn the
6 meeting.
7 (Whereupon the meeting was adjourned at
8 6:55 p.m.)
End, Page 471

I note that prior to Bajorek’s reference to your ruling, Consultant Wallis remarked, CONSULTANT WALLIS: But if you want a
5 comment, it looked as if there could be a significant
6 point here, I mean it's something that is not trivial
7 to look at and see is there a question here and what's
8 the evidence for --

And following the presentations by Mark Leyse and Robert Leyse, the Chairman of the Committee stated:
Page 191 of the transcript.
CHAIR BANERJEE: Thank you both very much.
16 We appreciate your comments, and we will certainly
17 take them into account.

However, following the above first cited remark by Bajorek, the ACRS has dropped all discussion of the Leyse submittals. And also,

19 MEMBER ABDEL-KHALIK: And I think from the
20 committee's perspective, we await the staff's
21 evaluation and we will review the staff's evaluation.

Although Bajorek as well as the members of the subcommittee were on the scene during the Leyse presentations, they made no remarks while the Leyses were on the bridge line. Instead, the remarks by Bajorek and the subcommittee came at the end of the meeting when neither Leyse could respond.

In closing, I repeat, please fill me in on the details of communications between Bajorek and perhaps others and yourself and/or your staff regarding the timing of the NRC’s evaluation of PRM-50-93.

Robert H. Leyse bobleyse@aol.com

I did not send this to ACRS

I need about 15 minutes to make a presentation via bridge line to the ACRS Planning and Procedures Subcommittee that will hold a meeting on February 3, 2010. I will be describing the need for ACRS procedures that will at least frown upon the unprofessional treatment that my partner and I encountered following our presentation to the ACRS SUBCOMMITTEE ON THERMAL HYDRAULIC PHENOMENA, MONDAY, OCTOBER 18, 2010 (ML103190234).

Briefly, we had 10 minutes for a succinct presentation at about 1:00 PM at the meeting. The subcommittee members had no comments at that time, and the meeting moved on at about 1:10 PM as the Chairman thanked us:
CHAIR BANERJEE: Thank you both very much.
16 We appreciate your comments, and we will certainly
17 take them into account.

My partner and I were then removed from the capability of addressing the subcommittee and we thought that any discussion of our presentation was ended.

The “trouble” began near the end of the meeting, sometime after 6:30 PM, page 467 of the transcript:
CONSULTANT KRESS: I found it very unusual
17 that public comments are made to the subcommittee.
18 Those usually go to the full committee.

As the discussion proceeded it became increasingly infuriating to the blocked Leyses.

The crowning blow came from Bajorek, pages 469 and 470 of the transcript:

MR. BAJOREK: He did make the point that
23 while there was a user need letter, point out and the
24 research was supposed to have responded by I think the
25 end of August. That was the original schedule. But
because they amended their own petition, and submitted
2 another petition, OGC decided to lump it together and
3 that window of time has moved out.
4 CHAIR BANERJEE: Okay. Well with that, I
5 think I'd like to thank you all and adjourn the
6 meeting.
7 (Whereupon the meeting was adjourned at
8 6:55 p.m.)

So, I need about 15 minutes to discuss the need for ACRS procedures that will avoid a repetition of this outrageous development. I will have no handouts; however, I will refer to appropriate parts of the transcript.

As an aside, but for the record, my belief is that with the exception of the Leyse presentations, the meeting was a waste of the taxpayers’ money.

ACRS Thermal Hydraulic Subcommittee and RBHT, October 18, 2010

Rod Bundle Heat Transfer at ACRS Subcommittee

RBHT was discussed by Seungjin Kim, Assistant Professor, Pennsylvania State University. His slides list six reports that were submitted to NRC during 2008. Kim’s list is in blue.
A corresponding list in black type is from Penn State University Reports as reported by NRC, McGinty to Leyse, April 16, 2010, (ML100950085). McGinty discloses that only one of the six reports is available to the public and it was not placed in ADAMS until 07/31/2010 (ML102290227). Three of Kim’s six reports have no publishing date set. Another is now predecisional but is expected to be published by December 2011. Finally, NRC expects to publish NUREG/CR 6975 as a public document by December 2010; however, it is not yet in ADAMS.

Click to enllarge the following:

ACRS wrote to NRC Chairman, "... several deficiences in the current regulations."

Bob Leyse Submitted the following to ACRS as noted. This is not a part of the attachments to the transcript of that meeting, however, it is in ADAMS, ML102530135.

ACRS Subcommittee on Plant License Renewal September 8, 2010, Room T–2B1, 11545 Rockville Pike, Rockville, Maryland. (Palo Verde)

Facts for the Subcommittee and for the record:

1. On May 23, 2007 Bonaca wrote Kline, SUBJECT: PROPOSED TECHNICAL BASIS FOR THE REVISION TO 10 CFR 50.46 LOCA EMBRITTLEMENT CRITERIA FOR FUEL CLADDING MATERIALS, ML071490090. Bonaca wrote, “The requirements of 10 CFR 50.46 (a) and (b) limit the amount of embrittlement that may occur as result of a design basis accident. They specify limits for the peak clad temperature, the global oxidation of cladding, and the local oxidation of cladding. There are several deficiencies with the current regulations. The correlation specified for the rates of steam reaction with the cladding is viewed by the technical community as an anachronism.” Now, Appendix K to Part 50--ECCS Evaluation Models, Item 5, specifies that the rate of energy release from the metal/water reaction shall be calculated using the Baker-Just equation and § 50.46 Acceptance Criteria, item (b)(1) specifies the peak clad temperature, 2200 degrees.

2. The NRC staff fiercely defends Baker-Just in its Technical Safety Analysis, ML041210109, April 29, 2004, “The Baker-Just correlation using the current range of parameter inputs is conservative and adequate to assess Appendix K ECCS performance. Virtually every data set published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F.”

3. The nuclear power industry fiercely defends Baker- Just in its Industry Comments, ML101040678, April 12, 2010, “The Baker-Just correlation, using the current range of parameter inputs, has been shown to be conservative and adequate to assess Appendix K ECCS performance. Data published since the Baker-Just correlation was developed has clearly demonstrated the conservatism of the correlation above 1800°F”

4. Contrary to the exceptionally firm consistency between the NEI and NRC appraisals of Baker-Just, the pertinent data sets published since the Baker-Just correlation was developed have clearly demonstrated the non-conservatism of the Baker-Just correlation above 1800°F. The NRC has not recognized that investigations that involve heating of single specimens of zirconium alloys in steam do not yield applicable data for the temperature or range of temperatures at which thermal runaway is initiated in LWRs.


5. NRC has apparently never studied Baker-Just (ML050550198) and until April 2010 it did not even have copies of the key references. Figure 16 is copied from page 37 of the Baker-Just report ML050550198.

Note to reader: Go to ML050550198, page 37 to view the Figure 16. This blog does not copy that.
Only the Lemmon data includes the pertinent temperature region. The Lemmon report, ML100570218, was not acquired by NRC until April, 2010. Thus, NRC never studied Baker-Just. Figure C-1 is from Lemmon page C-4; the adjacent figure is excerpted from the flow sheet, Figure C-3 on page C-5.

Another note to reader: Go to ML100570218, pages C-4 and C-5 to view
Figures C-1 and C-5.

Lemmon induction heated a zircaloy-2 cylinder, 2” long by 0.5” dia. in steam.
6. It is absurd to license the emergency cooling of tons of zirconium alloy having thousands of square feet of interfacial surface area based on the limited investigations that yielded the Baker-Just equation. Despite this, Appendix K to Part 50--ECCS Evaluation Models, Item 5, specifies that the rate of energy release from the metal/water reaction shall be calculated using the Baker-Just equation and § 50.46 Acceptance Criteria, item (b)(1) specifies the 2200 degrees.

7. Data from multi-rod (assembly) severe fuel damage experiments (e.g., the LOFT LP-FP-2 experiment) show the Baker-Just equation is non-conservative for calculating the temperature at which thermal runaway would occur in the event of a LOCA.

8. Investigations by P. Hofmann et al. at Forschungszentrum Karlsruhe reveal that the Baker-Just equation is non-conservative for calculating the temperature at which thermal runaway will occur in a LOCA. Their report is, Physico-Chemical Behavior of Zircaloy Fuel Rod Cladding Tubes During LWR Severe Accident Reflood, Part I: Experimental results of single rod quench experiments, FZKA 5846, http://bibliothek.fzk.de/zb/berichte/FZKA5846.pdf

On page 5 of 177: A series of separate-effects tests is being carried out on Zircaloy PWR fuel rod cladding to study the enhanced oxidation which can occur on quenching. In these tests, performed in the QUENCH rig, single tube specimens are heated by induction to a high temperature and then quenched by water or rapidly cooled down by steam injection.

On gage 12 of 177: No significant temperature excursion during quenching occurred such as had been observed for example in the quenched (flooded) CORA-bundle tests This absence of any temperature escalation is believed to be due to the high radiative heat losses in the QUENCH rig.

And in, “CORA Experiments on the Materials Behavior of LWR Fuel Rod Bundles at High Temperatures,” NUREG/CP-0119, Vol. 2, Proceedings of the Nineteenth Water Reactor Safety Information Meeting. ML042230460, P. Hofmann et al.

On page 98 of 493: The critical temperature above which uncontrolled temperature escalation takes place due to the exothermic zirconium/steam reaction crucially depends on the heat loss from the bundle; i.e., on bundle insulation. With the good bundle insulation in the CORA test facility, temperature escalation starts between 1100 and 1200°C (2012 to 2192°F), giving rise to a maximum heating rate of 15 K/sec.

9. It is amazing that the ACRS has never reviewed Baker-Just in the course of producing its recommendations regarding the initial licensing, the extended licensing and the licensing of power level increases of numerous American light water reactors.

Final note to reader: The following is copied from the from the FSAR for the Palo Verde Units, and this reveals the basis for licensing the ECCS at the Palo Verde units.
Palo Verde, Units 1, 2 and 3 - Updated Final Safety Analysis Report, Revision 14.
ML072250202
2007-06-30

PVNGS UPDATED FSAR
EMERGENCY CORE COOLING SYSTEM
June 2007 6.3-76 Revision 14

6.3.3 PERFORMANCE EVALUATION

6.3.3.1 Introduction and Summary

10 CFR 50.46 provides acceptance criteria for Emergency Core
Cooling Systems (ECCS) for light-water nuclear power reactors
[Reference 1]. The ECCS performance analyses described in this
section demonstrate that the PVNGS ECCS design satisfies these
criteria.

The PVNGS ECCS performance analyses encompass a wide range of
Reactor Coolant System (RCS) break locations and sizes, including
both large and small break Loss-of-Coolant Accident (LOCAs). The
limiting break, which results in the closest approach to 10 CFR
50.46 acceptance criterion for peak clad temperature, is a 0.6
DEG/PD (Double-Ended Guillotine in the Reactor Coolant Pump
Discharge leg) as noted in UFSAR Section 6.3.3.2. The limiting
break, which results in the closest approach to 10 CFR 50.46
acceptance criterion maximum clad oxidation (or local clad
oxidation), is a 0.8 DEG/PD as noted in UFSAR Section 6.3.3.2.
For these limiting breaks, the PVNGS ECCS design meets the
acceptance criteria of 10 CFR 50.46 as follows:

Criterion 1: Peak Cladding Temperature. ". . .The
calculated maximum fuel element cladding
temperature shall not exceed 2200 degrees F. . . ."
For the limiting break, the PVNGS ECCS
performance analysis yielded a peak cladding
temperature of 2110 degrees F.

PVNGS UPDATED FSAR
EMERGENCY CORE COOLING SYSTEM
June 2007 6.3-130 Revision 14

6.3.6 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50,
Section 50.46, "Acceptance Criteria for Emergency Core
Cooling Systems for Light Water Nuclear Power Reactors."