Sunday, December 5, 2010

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."

PRM-50-93 and PRM-50-95 at Full ACRS, December 2, 1010

Presentation to Full ACRS December 2, 2010

I’m Bob Leyse. This presentation is directed to two PRMs that were originated by Mark Leyse. These are PRM-50-93 and PRM-50-95, and today I am standing in for Mark. I’ll move through the 6 page handout in the allotted 5 minutes.

Moving to page 1 of the handout,

NRC should not authorize Plant License Renewals or Power Uprates prior to its resolution of PRM-50-93 and PRM-50-95.

The 2200 degree Fahrenheit PCT limit is too high. The 2200 PCT limit is based on embitterment criteria. The Baker-Just equation was placed into 50.46 and it has been convenient in licensing. Its current use in 50.46 is fiercely defended by the NRC.

Not in the handout, is an incorrect remark by Bajorek at the joint meeting of three ACRS subcommittees on May 31, 2002. “Note by the way Baker-Just in some of the earlier data was based on zirconium data only.” In fact, Baker-Just is very predominantly based on experiments with Zircaloy-2 at Bettis and Battelle. NRC did not even have those references until I appealed to the OGC and then the documents were acquired and placed in ADAMS. The reports are:

WAPD-104 ADAMS Accession No. ML100900446
BMI-1154 ADAMS Accession No. ML100570218

Go to page 2:

At another point in that joint meeting of the three ACRS subcommittees on May 31, 2002, we hear from

MEMBER WALLIS: 2200 has a very iffy
basis. The only justification really is that it is
worked over 30 or 40 years.

However, Member Wallis is wrong. There is nothing iffy about 2200. Very clearly, 2200 is too high and there is nothing iffy about that. Perhaps the most impressive evidence comes from experiment LOFT LP-FP-2 where thermal runaway of the fuel bundle was initiated in the 2060 to 2240 degree Fahrenheit range. And, the series of CORA experiments at Karlsruhe yielded thermal runaway over a range from about 1800 to 2200 degrees Fahrenheit. The CORA experiments used bundles of electrically heated rods having Zirconium alloy cladding and uranium dioxide fuel pellets.

On page 3 of the handout, note that Mark Leyse and Robert Leyse jointly made a 10 minute presentation to the ACRS Thermal Hydraulics Phenomena Subcommittee on Monday, October 18, 2010. Close to the end of the meeting the subcommittee briefly discussed the matter.

To save time, please skip to page 6,

Discussing the review of PRM-50-93 and PRM-50-95, at the October 18 meeting we hear from
MEMBER ABDEL-KHALIK: "And I think from the committee's perspective, we await the staff's evaluation and we will review the staff's evaluation."

Now, one primary mandate of the ACRS is; “to initiate reviews of specific generic matters or nuclear facility safety-related items.” In line with that mandate, I believe that ACRS should evaluate PRM-50-93 and PRM-50-95 in parallel and likely in advance of the NRC’s technical evaluations.

For emphasis I am repeating from page 1: NRC should not authorize Plant License Renewals or Power Uprates prior to its resolution of PRM-50-93 and PRM-50-95.

Thank you.

Handout for Bob Leyse Presentation to Full ACRS, 6 pages, December 2, 2010
Note: This handout to the Full ACRS is a copy of a comment that was submitted to the NRC by Bob Leyse regarding Mark Leyse Petitions for Rulemaking PRM-50-93 and PRM-50-95.

November 26, 2010

Annette L. Vietti-Cook
Secretary
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

Attention: Rulemakings and Adjudications Staff

COMMENTS ON PRM-50-93 AND PRM-50-95; NRC-2009-0554

NRC should not authorize Plant License Renewals or Power Uprates prior to its resolution of PRM-50-93 and PRM-50-95.

The 2200 degree Fahrenheit PCT limit is too high. The 2200 PCT limit is based on embitterment criteria. The Baker-Just equation was placed into 50.46 and it has been convenient in licensing. Its current use is fiercely defended by the NRC.

According to analyses funded by NRC, when the Baker-Just correlation is applied, the predicted thermal runaway starts at 2600 degrees Fahrenheit, while the alternative Cathcart-Pawel correlation of Reg. Guide 1.157 yields runaway at 2700. This is detailed on page 28 of PRM-50-93.

At a joint meeting of three ACRS subcommittees on May 31, 2002, there is the following pertinent exchange:

MR. LAUBEN: That's it. Sure. No.
That's an easy and quantifiable way to compare it. It
just gives you a minimum measure because what's really
true because of the slope changes so much is that you
can see a much bigger difference. In general I would
say I could never achieve turn-around much above 2300
in the limited 100 calculations I did with Baker-Just
but I could reach something as close to 2800 with
Cathcart-Pawel. Now that's –

MEMBER WALLIS: Maybe you need to show
these calculations. Something more convincing than
what we heard today --

At another point in that joint meeting of three ACRS subcommittees on May 31, 2002:

MEMBER WALLIS: 2200 has a very iffy
basis. The only justification really is that it is
worked over 30 or 40 years. If you are going to
change it you're going to have to have some really
good arguments.

However, Member Wallis is wrong. There is nothing “iffy” about 2200. At Karlsruhe it had already been clearly demonstrated that 2200 is too high and there is nothing “iffy” about the fact that 2200 is too high. An array of experiments having multirod assembles of rods with zirconium alloy cladding reveal that thermal runaway begins well below the 2600 to 2700 range. Perhaps the most impressive is LOFT LP-FP-2 where thermal runaway of the fuel bundle was initiated in the 2060 to 2240 degree Fahrenheit range. And, the series of CORA experiments at Karlsruhe with bundled electrically heated rods having Zirconium alloy cladding and uranium fuel pellets, yielded thermal runaway over a range from about 1800 to 2200 degrees Fahrenheit.

Although PRM-50-93 is dated November 2009, there is little evidence that the NRC has pursued its evaluation. On April 26, 2010, NRR issued a USER NEED REQUEST FOR TECHNICAL ANALYSIS OF PETITION FOR RULEMAKING ON 10 CFR 50.46 (PRM-50-93) and at that time the activity was (finally) assigned a high priority. Quoting from the User Need Request, "The requested deliverable for this user need is a technical letter report. Your office provided an outstanding technical analysis [reference 2] of a similar rulemaking petition, and we request the final deliverable for this user need be in this same format. We also request that a draft of your report be provided for comment by August 31, 2010 and the final report by September 30, 2010. We will provide comments on the draft within one week of receipt."

However: On October 27, 2010, the NRC published for public comment a notice of consolidation of petitions for rulemaking. The PRMs to be consolidated are PRM-50-93 filed by Mark Edward Leyse on November 17, 2009, and PRM-50-95 filed on June 7, 2010, by Mark Edward Leyse and Raymond Shadis, on behalf of the New England Coalition. What Mark Leyse filed on June 7, 2010 was not a PRM, it was a 2.206 petition. It appears that by consolidating these actions by Mark Leyse, the NRC has substantially extended the deadline for producing a Technical Letter Report regarding PRM-50-93. Nevertheless, the priority is established by the technical facts that are in the record and high priority attention by the NRC reviewers remains warranted.

In fact, Mark Edward Leyse first learned about the extended deadline when the ACRS Thermal Hydraulics Phenomena Subcommittee briefly discussed the matter on Monday, October 18, 2010. Mark Leyse and Robert Leyse had jointly made a 10 minute presentation, and at the end of the meeting the subcommittee discussed the matter as follows:

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.

Now, it is unlikely that the combined review of PRM-50-93 and PRM-50-95 adds sufficient complexity and data to justify a one year extension to the deadline for producing the Technical Analysis that is to be the basis of a recommendation to the NRC Commissioners for action on PRM-50-93 and PRM-50-95. Certainly, a substantial amount of review of PRM-50-93 should have been already completed prior to the merging of PRM-50-93 with the recent PRM-50-95.


Robert H. Leyse
P. O. Box 2850
Sun Valley, ID 83353

2200 degrees Fahrenheit is Too High

PRESENTATON BY BOB LEYSE TO ACRS SUBCOMMITTEE ON POWER UPRATE, NOVEMBER 17, 2010

2200 FAHRENHEIT IS TOO HIGH

I’m Bob Leyse and I have been in this business since 1950. I’ll race through the slide in 10 minutes. The slide covers two PRMs by Mark Leyse. The Committee is urged to digest those after this meeting, and that will take longer than 10 minutes, however, the members can certainly justify applying that time in their billing to ACRS. The slide has the ML numbers.

There are two items: the 2200 degree Fahrenheit PCT limit is too high and crud has a substantial impact on the PCT during a LOCA. Moving for a moment to today’s meeting, most of the AREVA presentation is reasonably not available to the public, however, I think it is likely that none of the KATHY games include the impact of a range of crud deposits.

OK, back to the slide. This is called the POWER UPRATE COMMITTEE, which presupposes that Power Uprates are in order. What we really need is a Power Level Review Committee.

The 2200 degree Fahrenheit PCT limit is too high. The 2200 PCT limit is based on embitterment criteria. The Baker-Just equation was placed into 50.46 and it has been convenient in licensing. According to analyses funded by NRC, when the Baker-Just correlation is applied, the predicted thermal runaway starts at 2600 degrees Fahrenheit, while the alternative Cathcart-Pawel correlation of Reg. Guide 1.157 yields runaway at 2700. However, an array of experiments having multirod assembles of rods with zirconium alloy cladding reveal that thermal runaway begins well below the 2600 to 2700 range. Perhaps the most impressive is LOFT LP-FP-2 where thermal runaway of the fuel bundle was initiated in the 2060 to 2240 degree Fahrenheit range. The series of CORA experiments at Karlsruhe with Zirconium alloy cladding of bundled electrically heated rods yielded thermal runaway over a range from about 1800 to 2200 degrees Fahrenheit.

The NRC staff is taking PRM-50-93 very seriously, and so should the ACRS.
The current User Need Request from NRR to RES is High Priority.
The requested deliverable for this user need is a technical letter report and the initial due date for a thoroughly researched final report was September 30, 2010.

However: On October 27, 2010, the NRC published for public comment a notice of consolidation of petitions for rulemaking. The PRMs to be consolidated are PRM-50-93filed by Mark Edward Leyse on November 17, 2009, and PRM-50-95 filed on June 7,
2010, by Mark Edward Leyse and Raymond Shadis, on behalf of the New England Coalition. What Mark Leyse filed on June 7, 2010 was not a
PRM, it was a 2.206 petition. It appears that by consolidating these actions by Mark Leyse, the NRC has extended the deadline for producing a Technical Letter Report regarding PRM-50-93. Nevertheless, the priority is established by the technical facts that are in the record and diligent and timely attention by the ACRS is most certainly called for under its mandate “to initiate reviews of specific generic matters or nuclear facility safety-related items.”

Moving to the impact of crud; PRM-50-84 details the impact of crud on the steady-state temperature distribution and stored energy in the fuel at the onset of a postulated LOCA.

Crud increases the operating fuel rod surface temperature and fuel rod stored energy. Crud decreases the overall heat transfer coefficient at the fuel rod. Crud adversely impacts the coolant flow distribution throughout the reactor core (fuel rod locations with heavier crud layers have less flow). Thus crud leads to substantial increases in the PCT during a LOCA.

In its Advance Notice of Proposed Rulemaking: Performance-Based ECCS Acceptance Criteria, July 29, 2009, NRC addresses PRM-84 as follows: In summary, to address the technical concerns related to crud in the PRM-50-84 petitioner’s request for rulemaking, the NRC is considering amending § 50.46 to specifically identify crud as a parameter to be considered in best-estimate and Appendix K to Part 50 ECCS evaluation models.

PRM-50-84 reports that EPRI will complete a program during 2008 that will “… determine the effect of tenacious crud on fuel surface heat transfer.” So far, I have found no open reporting of this.

AREVA and Westinghouse have brochures that describe ultrasonic fuel cleaning services. The recent Westinghouse brochure lists more than 12 LWRs that have used Ultrasonic Fuel Cleaning for crud removal from fuel elements. And from the AREVA brochure I quote, “AREVA NP offers patented Electric Power Research Institute (EPRI) Ultrasonic Fuel Cleaning (UFC) to prevent uneven crud deposits that can negatively affect fuel performance.”
Also interesting is a patent application: Chemical Enhancement of Ultrasonic Fuel Cleaning. Here are a few sentences (Only read the three sentences that are in bold.)
A method for cleaning an irradiated nuclear fuel assembly includes chemically enhancing a technique utilizing an apparatus including a housing adapted to engage a nuclear fuel assembly. A set of ultrasonic transducers is positioned on the housing to supply radially emanating omnidirectional ultrasonic energy to remove deposits from the nuclear fuel assembly. Any corrosion products remaining after ultrasonic fuel cleaning will have exposed surfaces that are susceptible to chemical dissolution.

The mechanical cleaning is effective, but it is not 100% efficient because corrosion products remain on the fuel assemblies. It is estimated that ultrasonic cleaning removes up to 80% of the total corrosion product inventory on the fuel

According to the subject method, chemical addition is localized to the water in the ultrasonic cleaning chamber rather than throughout the primary system, which minimizes the total liquid waste generated by orders of magnitude. Less aggressive chemistries can be selected that take advantage of the ultrasonic fuel cleaning environment. Only the fuel assemblies are exposed to the chemicals, so there is less chemical cleanup required for the vessel or ex-core piping. In certain embodiments, the chemical addition steps could be applied to selected high flux assemblies that have high corrosion deposition, while other fuel assemblies could be cleaned only ultrasonically.


The references by Mark Leyse and J. S. Lee that are listed at the end of the handout each disclose that crud significantly increases the local surface temperature of the cladding and the stored energy within the fuel.

NRR and RES are continuing their preparation of the Technical Letter Report that is to be the basis for a timely recommendation to the NRC Commissioners regarding the disposition of PRM-50-93. In the meantime, ACRS should not concur with any Power Uprate proposal until PRM-50-93 is resolved.

I have about two minutes left. I have worked at GE, Hanford and San Jose; Westinghouse, Monroeville; DuPont, Savannah River; Argonne, and the Nuclear Safety Analysis Center at EPRI. Elsewhere, during the 1970s, I invented, branded and marketed the RADCAL GAMMA THERMOMTER. GE Hitachi references my IEEE paper that describes the gamma thermometer that is central to their current licensing report, "Gamma Thermometer System for LPRM Calibration and Power Shape Monitoring." October 6, 2010. Accession Number ML102810320.

For emphasis, I repeat, you may tell the Full Committee to not concur with any Power Uprate proposal until PRM-50-93 is resolved.

Thank you.

SLIDE: PRESENTATION TO SUBCOMMITTEE ON POWER UPRATE

2200°F is nonconservative. PRM-50-93(ML093290250)

Petitioner (Mark Leyse) requests that NRC revise 10 C.F.R. § 50.46(b)(1) to require that the calculated maximum fuel element cladding temperature not exceed a limit based on data from multi-rod (assembly) severe fuel damage experiments.

Mark Leyse also authored and submitted on behalf of New England Coalition a 2.206 petition requesting that NRC order the lowering of LBPCT of VYNPS (ML101610121). NRC recently converted this to PRM-50-95; the public comment period is now open.

The User Need Request for PRM-50-93, NRR to RES, is High Priority. (ML100770117)

Multirod severe fuel damage experiments reveal that 2200°F is too high.

LOFT LP-FP-2 experiment at INL Runaway began at 2060°F – 2240°F
CORA experiments at Karlsruhe Runaway began at 1832°F - 2192°F
PHEBUS B9R-2 Runaway began at <2200°f>Impact of crud PRM-50-84 (ML070871368)

Petitioner (Mark Leyse) requests that NRC amend Appendix K to Part 50—ECCS Evaluation Models I(A)(1), The Initial Stored Energy in the Fuel, to require that the steady-state temperature distribution and stored energy in the fuel at the onset of a postulated LOCA be calculated by factoring in the role that the thermal resistance of crud and/or oxide layers on fuel cladding plays in increasing the stored energy in the fuel. This requirement also needs to apply to any NRC approved best-estimate ECCS evaluation models used in lieu of Appendix K calculations.

To address the technical concerns related to crud ... in PRM-50-84 ... the NRC is considering amending § 50.46 to specifically identify crud as a parameter to be considered in best-estimate and Appendix K to Part 50 ECCS evaluation models.
ANPR: Performance-Based ECCS Acceptance Criteria, 07/29/2009, ML091250132.

Ultrasonic Fuel Cleaning under 10 C.F.R. § 50.59: An Areva brochure, Ultrasonic Fuel Cleaning, was recently updated, 11/05/2010. The Westinghouse brochure, NS-FS-0085, April 2009, reports, Ultrasonic Fuel Cleaning has been used at the following plants: ANO, Callaway, Catawba, Ft. Calhoun, McGuire, Millstone, Quad Cities, Seabrook, South Texas 1 & 2, Vogtle 1 & 2, Vandellos, and Watts Bar.

Mark Leyse coauthored, “Considering the Thermal Resistance of Crud in LOCA Analysis,” ANS 2009 Winter Meeting, November 15-19, 2009, Washington, D. C.

J. S. Lee, et al., “Effects of Crud on the Fuel Rod Integrity in Steady-State and LB-LOCA Condition,” 2008 Water Reactor Fuel Performance Meeting, Seoul, Korea.

Tuesday, November 2, 2010

Fouling at the Experimental Boiling Water Reactor 1959

ANL "management" did not appreciate this disclosure. I had been back at ANL following military service and within six months I accomplished the following among other work. There were others who had written reports that included modeling of the thermal characteristics of the reactor core. However, with the severe fouling the real situation was far from the design documents that were the basis of the work by the "experts."

Eight pages. Click to enlarge and use your back arrow to get back here.





















Monday, October 25, 2010

Leyse gamma thermometer at work -- for the record ML102810320

It has been around for a long time. This is a recent release.

NEDO-33197-A, Revision 3, "Gamma Thermometer System for LPRM Calibration and Power Shape Monitoring."

Accession Number ML102810320

Date 10/06/2010

Below is the cover page of the recent GE report and the Leyse reference. Click to enlarge and use your return arrow to get back here.


Thursday, October 21, 2010

User Need Request regarding PRM-50-93 from NRC's NRR to RES: ML100770117

Mark Edward Leyse summarized the User Need Request in his presentation to the ACRS Subcommittee on Thermal-Hydraulic Phenomena, October 18, 2010.

PRM-50-93 is the subject of a user need request, dated April 26, 2010, from Eric Leeds, Director, Office of Nuclear Reactor Regulation, to Brian Sheron, Director Office of Nuclear Reactor Research. NRR's user need request states that I cite extensive data from numerous multi-rod experiments and that their request is a high priority, with a target date of September 30, 2010.

The complete User Need Request follows: (click to enlarge and use your back button to rturn)









Sunday, October 3, 2010

Grandjean and LOCA ML081680167

Followiong is reference material copied from:


"A State-of-the-Art Review of Past Programmes Devoted to Fuel Behaviour Under Loss-of-Coolant Conditions. Part 3. Cladding Oxidation. Resistance to Quench and Post-Quench Loads".
ML081680167 2008-06-13

Resistance to Quench
and Post-Quench Loads.
Claude GRANDJEAN, Georges HACHE
DPAM/SEMCA 2008-093
DIRECTION DE LA PREVENTION DES ACCIDENTS MAJEURS
Service d'Etudes et de Modelisation du Combustible en Situations Accidentelles


3 ZIRCALOY OXIDATION (EXPERIMENTAL RESULTS AND MODELS SUBSEQUENT TO THE
1973 ECCS HEARING)..................................................................................................................................... 47
3.1 ISOTHERMAL OXIDATION KINETICS OF ZIRCALOY.................................................................................. 47
3.1.1 Tests by Biederman et al. (Worcester Polytechnic Institute, USA) ............................................... 47
3.1.2 Tests by Westerman & Hesson (Battelle Pacific Northwest Laboratories, USA).......................... 49
3.1.3 Tests by Suzuki and Kawasaki (JAERI, Japan) ............................................................................. 50
3.1.4 Tests by Cathcart, Pawel et al. (Oak Ridge National Laboratory, USA) ...................................... 51
3.1.4.1 Isothermal tests to determine oxidation kinetics ....................................................................... 52
3.1.4.2 Hydrogen absorption in one-sided oxidation tests..................................................................... 54
3.1.5 Tests by Brown and Healey (CEGB, UK) ..................................................................................... 56
3.1.6 Tests by Urbanic and Heidrick (AECL, Canada).......................................................................... 58
3.1.7 Tests by Leistikow, Schanz et al. (FZK, Germany)........................................................................ 62
3.1.8 Ocken report (EPRI, USA) ............................................................................................................ 69
3.1.9 Tests by Prater and Courtright (PNL, Richland, USA) ................................................................. 72
3.1.10 Tests by Moalem and Olander (Univ. Ca., Berkeley, USA) .......................................................... 75


3.2 EFFECT OF THE FINITE SIZE OF SAMPLES................................................................................................. 79
3.3 TRANSIENT OXIDATION......................................................................................................................... 81
3.3.1 Tests by Cathcart, Pawel et al....................................................................................................... 81
3.3.2 Tests by Leistikow, Schanz et al. ................................................................................................... 84
3.4 INFLUENCE OF INITIAL OXIDATION......................................................................................................... 90
3.4.1 TAGCIS/ TAGCIR tests ................................................................................................................. 90
3.4.2 Tests by Leistikow and Schanz ...................................................................................................... 91
3.5 EFFECT OF IRRADIATION AND/OR INITIAL HYDROGEN CONTENT ............................................................ 94
3.5.1 TAGCIR tests................................................................................................................................ 94
3.5.2 Contribution of CODAZIR test results .......................................................................................... 95
3.5.3 HYDRAZIR oxidation tests............................................................................................................ 97
3.5.3.1 Analysis of the 1997 test series ................................................................................................. 97
3.5.3.2 Analysis of the 1999 test series ................................................................................................. 99
3.5.3.3 Global analysis of all oxidation tests......................................................................................... 99
3.5.4 Contribution of CINOG test results............................................................................................. 101
3.5.5 Information provided by foreign test program results ................................................................ 102
3.5.5.1 JAERI test results .................................................................................................................... 102
3.5.5.2 ANL test results....................................................................................................................... 103
3.5.5.3 Various results........................................................................................................................ 103
3.5.6 Conclusions regarding the effects of irradiation and pre-hydriding........................................... 104


3.6 OXIDATION UNDER LIMITED STEAM SUPPLY – EFFECT OF HYDROGEN CONCENTRATION IN AN OXIDISING
ATMOSPHERE .................................................................................................................................................. 106
3.6.1 Abnormal oxidation and hydrogen absorption in tests by Chung and Kassner .......................... 106
3.6.2 Tests by Chung and Thomas (ANL)............................................................................................. 108
3.6.3 Tests by Uetsuka (KfK)................................................................................................................ 110
3.6.4 Tests by Furuta and Kawasaki (JAERI) ...................................................................................... 113
3.6.5 Tests by Uetsuka and Otomo (JAERI) ......................................................................................... 116
3.6.6 Tests by Prater and Courtright ................................................................................................... 117
3.6.7 Tests by Moalem and Olander..................................................................................................... 118
3.6.8 Conclusion: oxidation under restricted steam flow and diluted steam atmosphere.................... 119
3.7 OXIDATION AT HIGH PRESSURE............................................................................................................ 120
3.7.1 Tests by Pawel et al. .................................................................................................................... 120
3.7.2 Tests by Bramwell et al. .............................................................................................................. 123
3.7.3 Tests by Park et al. ...................................................................................................................... 126
3.7.4 Tests by Vrtilkova et al. ............................................................................................................... 128
3.7.5 Conclusions on oxidation at high pressure ................................................................................. 130

Thursday, August 26, 2010

Emirates Jaczko and more



So, this is now on the NRC's web site , it has been there since August 24 or thereabouts. I'm trying to find out what is going on, more later.


Monday, August 23, 2010

The Generic Issue Racket

Here is another item that I do not want to forget. In general, the NRC says that if it applies to all, it applies to none; certainly it applies to none on a timely basis.

Sunday, August 22, 2010

The Regulatory Guide Racket

I did not want to forget this entry. I'm thinking about the details while I collect references.

Saturday, August 21, 2010

Power level uprates are based on an anachronism


Following is excerpted from a 2007 ACRS letter, ML071490090. It refers to Baker-Just as an anachronism although it does not say Baker-Just. This is so very interesting because ACRS approved several power level uprates that were justified based on compliance with Baker-Just.

UNITED STATES NUCLEAR REGULATORY COMMISSION

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

WASHINGTON, DC 20555 - 0001

May 23, 2007

The Honorable Dale E. Klein
Chairman
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

SUBJECT: PROPOSED TECHNICAL BASIS FOR THE REVISION TO 10 CFR 50.46 LOCA EMBRITTLEMENT CRITERIA FOR FUEL CLADDING MATERIALS

DISCUSSION
Zirconium alloy cladding used in current power reactors is embrittled by hydrogen absorption during normal operation and by oxidation and absorption of oxygen during the temperature transient associated with a LOCA. 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. Cladding oxidation resulting from normal operation also contributes to embrittlement during a LOCA and can be significant in high-burnup fuel. Currently, not all licensees account for oxidation during normal operation in LOCA analyses. Also, since current requirements refer only to Zircaloy and ZIRLO cladding, the use of modern cladding alloys with superior performance requires regulatory exemptions.

Furthermore, in addition to its convenience in licensing power level increases, Baker-Just is referenced by power reactor licensees in their justification for pilot testing of "modern" cladding alloys. For example San Onofre's application to run some pilot M-5 assemblies is justified by Baker-Just. The approval by NRC is dated December 2009, over 2.5 years following the ACRS description of Baker-Just as an anachronism. Thus, San Onofre Units 2 and 3 will run some pilot M-5 bundles. They cite data reported by Framatome Cogema Fuels (FCF) to establish that M-5 metal-water reaction characteristics are bounded by Baker-Just. Click on the slide below to enlarge and then use your return arrow to get back here.

Sunday, August 1, 2010

Thermal Runaway and Baker-Just

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.

So, how did this fall through the cracks?

The Baker-Just equation was published by the AEC during 1962; long before NRC was established during January 1975. Although it has a very weak foundation, the Baker-Just equation has survived to this day because it is convenient for licensing of nuclear power plants. This equation was first applied to the licensing of nuclear power plants during the era of the AEC when “regulators” were very strongly encouraged to expedite the role of what was then called atomic energy. The AEC, under the direction of the Congressional Joint Committee on Atomic Energy, worked very closely with the industry’s lobbying group then called the Atomic Industrial Forum. The lobbying group for today’s nuclear power industry is named the Nuclear Energy Institute (NEI).

It is unlikely that the nuclear power industry has ever commissioned a study of the roots of the Baker-Just correlation. However, today’s nuclear power industry fiercely defends Baker- Just. This defense is well documented by the NEI in its very recent comment 16 opposing PRM-50-93, “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,” refer to ML101040678, Industry Comments on Petition for Rulemaking (PRM-50-93); Multi-Rod (Assembly) Severe Fuel Damage Experiments. Docket ID NRC-2009-0554, April 12, 2010.

The NRC also fiercely defends Baker-Just. In its analysis of PRM-50-76, “The Baker-Just correlation (Reference 4) 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.” refer to Memo to Matthews/Black-Technical Safety Analysis of PRM-50-76, A Petition for Rulemaking to Amend Appendix K to 10 CFR Part 50 and Regulatory Guide 1.157 - ML041210109, April 29, 2004.


Contrary to the exceptionally firm consistency between the NEI and NRC
evaluations 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 never admitted 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. Certainly, the NRC evaluators who produced ML041210109, April 29, 2004, should have been aware of the many references that are cited in PRM-50-93.




And, the following Letter from NRC to your blogger, ML100950085, documents in detail how Baker-Just fell through the cracks. NRC did not even have the key references in its files until your blogger persisted in demanding the key references. Click on the following to enlarge and use your back arrow to return here.







So, the above letter reveals:

  • that your blogger pursued the roots of Baker-Just,
  • that NRC did not believe that those roots constituted a significant part of its basis for denying PRM-50-76,
  • that those roots were not available in NRC files when it evaluated PRM-50-76,
  • that those roots were not transferred to NRC when it was created in 1974,
  • that NRC finally acquired and placed those roots in ADAMS during April 2010,
  • that NRC promptly informed your blogger when those roots became available in ADAMS.

It is interesting that NRC did not believe that those roots constituted a significant part of its basis for denying PRM-50-76 even though those roots were not then available in NRC files.

Furthermore: I'm adding the following on August 8, 2009.

When the NRC placed document BMI-1154 into ADAMS during April 2009 the document was incomplete. I recently found this out when I was seeking the references on page C-48. I contacted BMI at Pacific Northwest Laboratories (PNL) and I was told that there was no page C-48, "No, it does not have page C-48." So I copied and e-mailed the table of contents and I replied, "I wonder how that vital list of references vaporized." Next PNL told me on 8/4/2010, "Bob, Try this version. I found it on Energy Citations Database (www.osti.gov/energy/citations). It has C-48. Thanks. Cheryl Wiborg." I sent that information to the NRC and the complete copy was placed into ADAMS. NRC thanked me on 8/5/2010 as follows: "The NRC version in ADAMS has been updated (ML100570218). It may take a couple of days to copy to the NRC public server, but it has been fixed. Thanks for locating a complete copy of BMI-1154. John Boska, Indian Point Project Manager, NRR/DORL, U.S. Nuclear Regulatory Commission." And here is the latest in ADAMS, it has 125 pages, prior to the correction by Boska it had only 98 pages.

Report BMI-1154, "Studies Relating to the Reaction Between Zirconium & Water at High Temperatures." ML100570218 1957-01-03 125

So, I'll now add item 7 to the prior bulleted list because the above paragraph reveals:

  • that your blogger studied BMI-1154 and effected corrections, although he did not immediately find the omissions in the ADAMS copy.

Monday, July 19, 2010

The Germans, among others, have killed Baker-Just.

The Germans, among others, have killed Baker-Just.

Investigations by P. Hofmann and V. Noak at Forschungszentrum Karlsruhe reveal that the Baker-Just equation is non-conservative for calculating the temperature at which runaway oxidation 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, Institut für Materialforschung, Projekt Nukleare Sicherheitsforschung, Mai 1997. They report:

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.

No significant temperature excursion during quenching occurred such as had been observed for example in the quenched (flooded) CORA-bundle tests /4, 5/. 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,” P. Hofmann, S. Hagen, G. Schanz, G. Schumacher, L. Sepold, report:

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.

Mark Edward Leyse has initiated the following:

Mark E. Leyse 2.206 Petition to Lower the Licensing Basis Peak Cladding Temperature of Vermont Yankee in Order to Provide Necessary Margin of Safety in Event of LOCA ML101610121

Footnote 8, page 8 of the above 2.206 Petition is as follows:

Data from multi-rod (assembly) severe fuel damage experiments (e.g., the LOFT LP-FP-2 experiment) indicates that the Baker-Just and Cathcart-Pawel equations are both non-conservative for calculating the temperature at which an autocatalytic (runaway) oxidation reaction of Zircaloy would occur in the event of a LOCA. This, in turn, indicates that the Baker-Just and Cathcart-Pawel equations are both non-conservative for calculating the metal-water reaction rates that would occur in the event of a LOCA.

From the CONCLUSION, page 67:

Petitioner requests that the NRC order the licensee of VYNPS to lower the LBPCT of VYNPS in order to provide a necessary margin of safety—to help prevent a partial or complete meltdown—in the event of a LOCA. Experimental data indicates that VYNPS’s LBPCT of 1960°F does not provide a necessary margin of safety—to help prevent a partial or complete meltdown—in the event of a LOCA. Such data indicates that VYNPS’s LBPCT must be decreased to a temperature lower than 1832°F in order to provide a necessary margin of safety.

NRC Chairman Jaczko, Baker-Just, PRM-50-76

Jaczko visited the Vermont Yankee Nuclear Power Plant and vicinity on or about July 14, 2010. He should have stayed home and tended to business. All this is covered in the following five slides. Click to enlarge and use your back arrow to return for the next slide.





Tuesday, July 6, 2010

Sunday, July 4, 2010

BAKER-JUST and SPARKLERS (Single Rods ain’t Bundles)

I've discussed the 2200 degree Fahrenheit game that NRC plays; a game that is derived from the single rod tests that are referenced by Baker-Just. With bundle tests, runaway starts at much lower temperatures.

Analogies have weaknesses. However, the following from the Wall Street Journal is interesting. A bundle of sparklers yields wild behavior.


Friday, June 11, 2010

Blowout Protection and 2200 Fahrenheit

I've already posted a lot on this matter.

Following is an e-mail I sent to each of the great Commissioners of the NRC on June 10, 2010.

Rescinding the Denial of PRM-50-76

To the Commissioners:

You are urged to direct the NRC staff to promptly write a document for your approval that rescinds the denial of PRM-50-76 that was published in the Federal Register on September 6, 2005 (NRC-2002-0019-00130). The attachment to this e-mail, Rebuttal SECY-05-0113 Leyse R, has Leyse’s rebuttal in 14 point type interspersed within the NRC’s draft notice, SECY-03-0113, dated June 29, 2005.

Rescinding the denial of PRM-50-76 is vital. Under current regulations at 1OCFR50.46(b)(1) and Appendix K to 1OCFR Part 50, our nuclear power plants lack sufficient measures to prevent runaway to meltdown when blowout occurs during a loss-of-coolant accident. The Baker-Just equation does not accurately and conservatively indicate the conditions under which an autocatalytic (runaway) oxidation reaction of Zircaloy begins.

Reference: Appendix K to Part 50--ECCS Evaluation Models

I. Required and Acceptable Features of the Evaluation Models

5. Metal--Water Reaction Rate. The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation (Baker, L., Just, L.C., "Studies of Metal Water Reactions at High Temperatures, III. Experimental and Theoretical Studies of the Zirconium-Water Reaction," ANL-6548, page 7, May 1962).

Thursday, June 3, 2010

Draft Final Rule 50.46a Note 2200 Fahrenheit

§ 50.46a DRAFT FINAL RULE LANGUAGE

Risk-Informed Changes to Loss-of-Coolant Accident Technical Requirements
(ADAMS Accession no. ML101250271)
In August 2009, the Nuclear Regulatory Commission (NRC) published for public comment a
supplemental proposed § 50.46a risk-informed emergency core cooling system (ECCS) rule
(74 FR 40006). The comment period ended on January 22, 2010. The NRC has evaluated the
public comments and has prepared draft final rule language. To facilitate public involvement in
this rulemaking, the NRC is making this draft rule language public. The NRC has tentative
plans to hold a public meeting in early June 2010 to discuss resolution of public comments and
the associated draft final rule language. In the future, if significant changes are made to this
rule, the NRC may periodically update the publicly available rule language.
NOTE: The availability of this draft rule language is intended to inform
stakeholders of the current status of the NRC’s activities to
provide a risk-informed alternative to the current ECCS
requirements. This draft rule language may be incomplete or
in error in one or more respects and may be subject to further
revisions during the rulemaking process. The NRC is not
requesting formal public comments on this draft rule
language.
Any questions on the requirements may be addressed to the NRC rulemaking project manager,
Richard Dudley (301-415-1116; richard.dudley@nrc.gov).
1 Revised: April 30, 2010
1 List of Subjects
2
3
4 10 CFR Part 50
5 Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental
6 relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria,
7 Reporting and recordkeeping requirements.
8 10 CFR Part 52
9 Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early
10 site permit, Emergency planning, Fees, Inspection, Limited work authorization, Nuclear power
11 plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of
12 site, Reporting and recordkeeping requirements, Standard design, Standard design certification.
13 For the reasons set out in the preamble and under the authority of the Atomic Energy
14 Act of 1954, as amended; the Energy Reorganization Act of 1974; and 5 U.S.C. 553; the NRC is
15 adopting the following amendments to Title 10 of the Code of Federal Regulations (10 CFR)
16 Parts 50 and 52.
17 PART 50 -- DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES
18 1. The authority citation for part 50 continues to read as follows:
19 Authority: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat. 936, 937, 938,
20 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2132,
21 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88
22 Stat. 1242, as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750
23 (44 U.S.C. 3504 note); Energy policy Act of 2005, Pub. L. No. 109-58, 119 Stat. 194 (2005).
24 Section 50.7 also issued under Pub. L. 95-601, sec. 10, 92 Stat. 2951 as amended by Pub. L.
25 102-486, sec. 2902, 106 Stat. 3123 (42 U.S.C. 5841). Section 50.10 also issued under secs.
26 101, 185, 68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec. 102, Pub. L. 91-190, 83 Stat.
2 Revised: April 30, 2010
853 (42 U.S.C. 4332). 1 Sections 50.13, 50.54(dd), and 50.103 also issued under sec. 108, 68
2 Stat. 939, as amended (42 U.S.C. 2138).
3 Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42
4 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L. 91-
5 190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204, 88
6 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91, and 50.92 also issued under Pub. L. 97-
7 415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42
8 U.S.C. 2152). Sections 50.80 - 50.81 also issued under sec. 184, 68 Stat. 954, as amended (42
9 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237)
10 2. In § 50.34, paragraphs (a)(4) and (b)(4) are revised to read as follows:
11 § 50.34 Contents of application; technical information.
12
13 (a) * * *
14
15 (4) A preliminary analysis and evaluation of the design and performance of structures,
16 systems, and components of the facility with the objective of assessing the risk to public health
17 and safety resulting from operation of the facility and including determination of the margins of
18 safety during normal operations and transient conditions anticipated during the life of the facility,
19 and the adequacy of structures, systems, and components provided for the prevention of
20 accidents and the mitigation of the consequences of accidents.
21 (i) Analysis and evaluation of emergency core cooling system (ECCS) cooling
22 performance and the need for high point vents following postulated loss-of-coolant accidents
23 must be performed under the requirements of either § 50.46 or § 50.46a, and § 50.46b for
24 facilities whose operating licenses were issued after December 28, 1974, but before
25 [EFFECTIVE DATE OF RULE], and for facilities for which
3 Revised: April 30, 2010
construction permits may be issued after [EFFECTIVE DAT 1 E OF RULE] and are demonstrated
2 under § 50.46a(c)(2) to have designs that are similar to the designs of reactors licensed before
3 [EFFECTIVE DATE OF RULE].
4 (ii) Analysis and evaluation of ECCS cooling performance and the need for high point
5 vents following postulated loss-of-coolant accidents (LOCAs) must be performed under the
6 requirements of § 50.46 and § 50.46b for facilities for which construction permits may be issued
7 after [EFFECTIVE DATE OF RULE] and are not demonstrated under § 50.46a(c)(2) to have
8 designs that are similar to the designs of reactors licensed before [EFFECTIVE DATE OF
9 RULE].
10 * * * * *
11
12 (b) * * *
13
14 (4) A final analysis and evaluation of the design and performance of structures, systems,
15 and components with the objective stated in paragraph (a)(4) of this section and taking into
16 account any pertinent information developed since the submittal of the preliminary safety
17 analysis report.
18 (i) Analysis and evaluation of ECCS cooling performance following postulated LOCAs
19 must be performed under the requirements of either § 50.46 or § 50.46a, and § 50.46b for
20 facilities whose operating licenses were issued after December 28, 1974, but before
21 [EFFECTIVE DATE OF RULE], and for facilities whose operating licenses are issued after
22 [EFFECTIVE DATE OF RULE] and are demonstrated under § 50.46a(c)(2) to have designs that
23 are similar to the designs of reactors licensed before [EFFECTIVE DATE OF RULE].
24 (ii) Analysis and evaluation of ECCS cooling performance following postulated LOCAs
25
4 Revised: April 30, 2010
must be performed under the requirements of 1 §§ 50.46 and 50.46b for facilities whose operating
2 licenses are issued after [EFFECTIVE DATE OF RULE] and are not demonstrated under
3 § 50.46a(c)(2) to have designs that are similar to the designs of reactors licensed before
4 [EFFECTIVE DATE OF RULE].
5
6 * * * * *
7
8 3. In § 50.46, paragraph (a) is amended by adding an introductory paragraph and
9 revising paragraph (a)(1)(i) to read as follows:
10 § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear
11 power plants.
12 (a) Each boiling or pressurized light-water nuclear power reactor fueled with uranium
13 oxide pellets within cylindrical zircalloy or ZIRLO cladding must be provided with an ECCS. The
14 ECCS system must be designed under the requirements of this section or § 50.46a for facilities
15 whose operating licenses were issued before [EFFECTIVE DATE OF RULE]; for facilities
16 whose operating licenses, combined licenses under part 52 of this chapter, or manufacturing
17 licenses under part 52 of this chapter are issued after [EFFECTIVE DATE OF RULE] and are
18 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
19 licensed before [EFFECTIVE DATE OF RULE]; and for design approvals and design
20 certifications under part 52 of this chapter issued after [EFFECTIVE DATE OF RULE] that are
21 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
22 licensed before [EFFECTIVE DATE OF RULE]. The ECCS system must be designed under the
23 requirements of this section for facilities whose operating licenses, combined licenses under
24 part 52 of this chapter, or manufacturing licenses under part 52 of this chapter are issued after
25
5 Revised: April 30, 2010
1 [EFFECTIVE DATE OF RULE] and are not
2 demonstrated under § 50.46a(c)(2) to have designs that are similar to the designs of reactors
3 licensed before [EFFECTIVE DATE OF RULE]; and for design approvals and design
4 certifications under part 52 of this chapter that are not demonstrated under § 50.46a(c)(2) to
5 have designs that are similar to the designs of reactors licensed before [EFFECTIVE DATE OF
6 RULE].
7 (1)(i) The ECCS system must be designed so that its calculated cooling performance
8 following postulated LOCAs conforms to the criteria set forth in paragraph (b) of this section.
9 ECCS cooling performance must be calculated in accordance with an acceptable evaluation
10 model and must be calculated for a number of postulated LOCAs of different sizes, locations,
11 and other properties sufficient to provide assurance that the most severe postulated LOCAs are
12 calculated. Except as provided in paragraph (a)(1)(ii) of this section, the evaluation model must
13 include sufficient supporting justification to show that the analytical technique realistically
14 describes the behavior of the reactor system during a LOCA. Comparisons to applicable
15 experimental data must be made and uncertainties in the analysis method and inputs must be
16 identified and assessed so that the uncertainty in the calculated results can be estimated. This
17 uncertainty must be accounted for, so that, when the calculated ECCS cooling performance is
18 compared to the criteria set forth in paragraph (b) of this section, there is a high level of
19 probability that the criteria would not be exceeded. Appendix K, Part II Required
20 Documentation, sets forth the documentation requirements for each evaluation model. This
21 section does not apply to a nuclear power reactor facility for which the certifications required
22 under § 50.82(a)(1) have been submitted.
23
24 * * * * *
25
6 Revised: April 30, 2010
4. Section 50.46a 1 is redesignated as § 50.46b, and a new § 50.46a is added to read as
2 follows:
3 § 50.46a Alternative acceptance criteria for emergency core cooling systems for light4
water nuclear power reactors.
5 (a) Definitions. For the purposes of this section:
6 (1) Changes enabled by this section means changes to the facility, technical
7 specifications, and procedures that satisfy the alternative ECCS analysis requirements under
8 this section but do not satisfy the ECCS requirements under 10 CFR 50.46.
9 (2) Evaluation model means the calculational framework for evaluating the behavior of
10 the reactor system during a postulated design-basis loss-of-coolant accident (LOCA). It
11 includes one or more computer programs and all other information necessary for application of
12 the calculational framework to a specific LOCA, such as mathematical models used,
13 assumptions included in the programs, procedure for treating the program input and output
14 information, specification of those portions of analysis not included in computer programs,
15 values of parameters, and all other information necessary to specify the calculational procedure.
16 (3) Loss-of-coolant accidents (LOCAs) means the hypothetical accidents that would
17 result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant
18 makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and
19 including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor
20 coolant system. LOCAs involving breaks at or below the transition break size (TBS) are
21 design-basis accidents. LOCAs involving breaks larger than the TBS are beyond design-basis
22 accidents.
23 (4) Operating configuration means those plant characteristics, such as power level,
24 equipment unavailability (including unavailability caused by corrective and preventive
25
7 Revised: April 30, 2010
maintenance), and equipment capabilit 1 y that affect plant response to a LOCA.
2 (5) Transition break size (TBS) for reactors licensed before [EFFECTIVE DATE OF
3 RULE] is a break area equal to the cross-sectional flow area of the inside diameter of the largest
4 piping attached to the reactor coolant system for a pressurized water reactor, or the inside
5 diameter of the larger of the feedwater line inside containment or the residual heat removal line
6 inside containment for a boiling water reactor. For reactors licensed after [EFFECTIVE DATE
7 OF RULE], and for design certifications, design approvals, and and manufacturing licenses
8 approved or issued after [EFFECTIVE DATE OF RULE], the TBS will be determined on a plant9
specific basis.
10 (b) Applicability and scope.
11 (1) The requirements of this section may be applied to each boiling or pressurized
12 light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical zircalloy or
13 ZIRLO cladding whose operating license was issued prior to [EFFECTIVE DATE OF RULE]; to
14 each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets
15 within cylindrical zircalloy or ZIRLO cladding whose operating license, combined license under
16 part 52 of this chapter or manufacturing license under part 52 of this chapter is issued after
17 [EFFECTIVE DATE OF RULE] and whose design is demonstrated under § 50.46a(c)(2) to be
18 similar to the designs of reactors licensed before [EFFECTIVE DATE OF RULE]; and to each
19 boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within
20 cylindrical zircalloy or ZIRLO cladding whose design approval or design certification under part
21 52 of this chapter is demonstrated under § 50.46a(c)(2) to be similar to the designs of reactors
22 licensed before [EFFECTIVE DATE OF RULE]. The requirements of this section do not apply
23 to a reactor for which the certification required under § 50.82(a)(1) has been submitted.
24 (2) The requirements of this section are in addition to any other requirements applicable
8 Revised: April 30, 2010
to ECCS set forth in this part, with the exception of § 50.46. The 1 criteria set forth in paragraphs
2 (e)(3) and (e)(4) of this section, with cooling performance calculated in accordance with an
3 acceptable evaluation model or analysis method under paragraphs (e)(1) and (e)(2) of this
4 section, are in implementation of the general requirements with respect to ECCS cooling
5 performance design set forth in this part, including in particular Criterion 35 of Appendix A to this
6 part.
7 (c) Application. (1) An applicant, permit holder, or licensee of a facility, or other entity
8 seeking to implement this section shall submit an application for a license amendment under
9 § 50.90 that contains the following information:
10 (i) A written evaluation demonstrating applicability of the results in NUREG-1829,
11 “Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the Elicitation Process”;
12 March 2008 and NUREG-1903, “Seismic Considerations for the Transition Break Size”;
13 February 2008” to the licensee’s facility. As part of this evaluation, the application must contain
14 a plant specific analysis demonstrating that the risk of seismically-induced LOCAs larger than
15 the TBS is comparable to or less than the seismically-induced LOCA risk associated with the
16 NUREG-1903 results.
17 (ii) Identification of the approved analysis method(s) for demonstrating compliance with
18 the ECCS criteria in paragraph (e) of this section.
19 (iii) A description of the risk-informed evaluation process used in evaluating whether
20 proposed changes to the facility meet the requirements in paragraphs (d)(5) and (f), of this
21 section.
22 (iv) An applicant, permit holder, or licensee of a facility or other entity who wishes to
23 make changes enabled by this section without prior NRC review and approval must submit for
24 NRC approval a process to be used for evaluating the acceptability of these changes; including:
9 Revised: April 30, 2010
(A) A description of the approach, methods, 1 and decision making process to be used for
2 evaluating compliance with the acceptance criteria in paragraphs (f)(1), (f)(2), and (f)(3) of this
3 section, and
4 (B) A description of the PRA model and non-PRA risk assessment methods to be used
5 for demonstrating compliance with paragraphs (f)(4) and (f)(5) of this section.
6 (v) A description of non safety equipment that is credited for demonstrating compliance
7 with the ECCS acceptance criteria in paragraph (e) of this section.
8 (2) An applicant for a construction permit, operating license, design approval, design
9 certification, manufacturing license, or combined license seeking to implement the requirements
10 of this section shall, in addition to the information required by paragraph (c)(1) of this section,
11 submit an analysis demonstrating why the proposed reactor design is similar to the designs of
12 reactors licensed before [EFFECTIVE DATE OF RULE] such that the provisions of this section
13 may properly apply. The analysis must also include a recommendation for an appropriate TBS
14 and a justification that the recommended TBS is consistent with the technical basis for this
15 section.
16 (3) Acceptance criteria. The NRC may approve an application to use this section if:
17 (i) The evaluation submitted under paragraph (c)(1)(i) of this section demonstrates that
18 the NUREG-1829 results are applicable to the facility, and the risk of seismically-induced
19 LOCAs larger than the TBS is comparable to or less than the seismically-induced LOCA risk
20 associated with the NUREG-1903 results;
21 (ii) The method(s) for demonstrating compliance with the ECCS acceptance criteria in
22 paragraphs (e)(3) and (e)(4) of this section meet the requirements in paragraphs (e)(1) and
23 (e)(2) of this section;
10 Revised: April 30, 2010
(iii) The risk-informed evaluation process 1 the licensee proposes to use for making
2 changes enabled by this section is adequate for determining whether the acceptance criteria in
3 paragraphs (d)(5) and (f) of this section, have been met; and
4 (iv) Non safety equipment that is credited for demonstrating compliance with the ECCS
5 acceptance criteria in paragraph (e) of this section is identified in plant Technical Specifications.
6 (v) For all applicants other than those holding operating licenses issued before
7 [EFFECTIVE DATE OF RULE], the proposed reactor design is similar to the designs of reactors
8 licensed before [EFFECTIVE DATE OF RULE] and the applicant’s proposed TBS is consistent
9 with the technical basis of this section.
10 (d) Requirements during operation. A licensee whose application under paragraph (c) of
11 this section is approved by the NRC shall comply with the following requirements as long as the
12 facility is subject to the requirements in this section until the licensee submits the certifications
13 required by § 50.82(a):
14 (1) The licensee shall maintain ECCS model(s) and/or analysis method(s) meeting the
15 requirements in paragraphs (e)(1) and (e)(2) of this section;
16 (2) The licensee shall have leak detection systems available at the facility and shall
17 implement actions as necessary to identify, monitor and quantify leakage to ensure that adverse
18 safety consequences do not result from primary pressure boundary leakage from piping and
19 components that are larger than the transition break size.
20 (3) A change enabled by this section must, in addition to meeting other applicable NRC
21 requirements, be evaluated by a risk-informed evaluation demonstrating that the acceptance
22 criteria in paragraph (f) of this section are met.
23 (4) The licensee shall periodically maintain and upgrade, as necessary, its risk
24 assessments to meet the requirements in paragraph (f)(4) and (f)(5) of this section. The
11 Revised: April 30, 2010
maintenance and upgrading 1 shall be consistent with NRC-endorsed consensus standards on
2 PRA and must be completed in a timely manner, but no less often than once every four years.
3 Based upon a re-evaluation of the risk assessments after the periodic maintenance and
4 upgrading are completed, the licensee shall take appropriate action to ensure that the
5 acceptance criteria in paragraph (f) of this section, as applicable, are met. The PRA
6 maintenance and upgrading required by this section, and any necessary changes to the facility,
7 technical specifications and procedures as a result of this re-evaluation, shall not be deemed to
8 be backfitting under any provision of this chapter.
9 (5) For LOCAs larger than the TBS, operation in a plant operating configuration not
10 demonstrated to meet the acceptance criteria in paragraph (e)(4) of this section may not exceed
11 a short time. A short time is either a total of fourteen (14) days in any 12 month period or an
12 NRC-approved alternative time proposed by the licensee that is commensurate with the
13 mitigation capability available and has been demonstrated to be acceptable by a risk-informed
14 evaluation of the configuration-specific risk, defense-in-depth, and safety margins.
15 (6) The licensee shall perform an evaluation to determine the effect of all planned facility
16 changes and shall not implement any facility change that would invalidate the evaluation
17 performed pursuant to § 50.46a(c)(1)(i) demonstrating the applicability to the licensee’s facility
18 of the generic results in NUREG-1829 and NUREG-1903.
19 (e) ECCS Performance. Each nuclear power reactor or nuclear power reactor design
20 subject to this section must be provided with an ECCS that must be designed so that its
21 calculated cooling performance following postulated LOCAs conforms to the criteria set forth in
22 this section. The evaluation models for LOCAs must meet the criteria in this paragraph, and
23 must be approved for use by the NRC. Appendix K, Part II, to 10 CFR Part 50, sets forth the
24 documentation requirements for evaluation models.
12 Revised: April 30, 2010
(1) ECCS evaluation f 1 or LOCAs involving breaks at or below the TBS. ECCS cooling
2 performance at or below the TBS must be calculated in accordance with an evaluation model
3 that meets the requirements of either section I to Appendix K of this part, or the following
4 requirements, and must demonstrate that the acceptance criteria in paragraph (e)(3) of this
5 section are satisfied. The evaluation model must be used for a number of postulated LOCAs of
6 different sizes, locations, and other properties sufficient to provide assurance that the most
7 severe postulated LOCAs involving breaks at or below the TBS are analyzed. The evaluation
8 model must include sufficient supporting justification to show that the analytical technique
9 realistically describes the behavior of the reactor system during a LOCA. Comparisons to
10 applicable experimental data must be made and uncertainties in the analysis method and inputs
11 must be identified and assessed so that the uncertainty in the calculated results can be
12 estimated. This uncertainty must be accounted for, so that when the calculated ECCS cooling
13 performance is compared to the criteria set forth in paragraph (e)(3) of this section, there is a
14 high level of probability that the criteria would not be exceeded.
15 (2) ECCS analyses for LOCAs involving breaks larger than the TBS. ECCS cooling
16 performance for LOCAs involving breaks larger than the TBS must be calculated in accordance
17 with an evaluation model that meets the requirements of either section I to Appendix K of this
18 part, or the following requirements, and must demonstrate that the acceptance criteria in
19 paragraph (e)(4) of this section are satisfied. The evaluation model must include sufficient
20 supporting justification to show that the analytical technique realistically describes the behavior
21 of the reactor system during a LOCA. Comparisons to applicable experimental data must be
22 made and uncertainties in the analysis method and inputs must be identified and assessed so
23 that the uncertainty in the calculated results can be estimated. This uncertainty must be
24 accounted for, so that when the calculated ECCS cooling performance is compared to the
13 Revised: April 30, 2010
criteria set forth in paragraph (e)(4) of this section, there is a 1 high level of probability that the
2 criteria would not be exceeded. The evaluation model must be used for a number of postulated
3 LOCAs of different sizes, locations, and other properties sufficient to provide assurance that the
4 most severe postulated LOCAs larger than the TBS up to the double-ended rupture of the
5 largest pipe in the reactor coolant system are analyzed. These calculations may take credit for
6 the availability of offsite power and do not require the assumption of a single failure. Realistic
7 initial conditions and availability of safety-related or non safety-related equipment may be
8 assumed if supported by plant-specific data or analysis, and provided that onsite power can be
9 readily provided through simple manual actions to equipment that is credited in the analysis.
10 (3) Acceptance criteria for LOCAs involving breaks at or below the TBS. The following
11 acceptance criteria must be used in determining the acceptability of ECCS cooling performance:
12 (i) Peak cladding temperature. The calculated maximum fuel element cladding
13 temperature must not exceed 2200°F.
14 (ii) Maximum cladding oxidation. The calculated total oxidation of the cladding must not
15 at any location exceed 0.17 times the total cladding thickness before oxidation. As used in this
16 paragraph, total oxidation means the total thickness of cladding metal that would be locally
17 converted to oxide if all the oxygen absorbed by and reacted with the cladding locally were
18 converted to stoichiometric zirconium dioxide. If cladding rupture is calculated to occur, the
19 inside surfaces of the cladding must be included in the oxidation, beginning at the calculated
20 time of rupture. Cladding thickness before oxidation means the radial distance from inside to
21 outside the cladding, after any calculated rupture or swelling has occurred but before significant
22 oxidation. Where the calculated conditions of transient pressure and temperature lead to a
23 prediction of cladding swelling, with or without cladding rupture, the unoxidized cladding
24 thickness must be defined as the cladding cross-sectional area, taken at a horizontal plane at
14 Revised: April 30, 2010
the elevation of the rupture, if it occurs, or at the elevation of the 1 highest cladding temperature if
2 no rupture is calculated to occur, divided by the average circumference at that elevation. For
3 ruptured cladding the circumference does not include the rupture opening.
4 (iii) Maximum hydrogen generation. The calculated total amount of hydrogen generated
5 from the chemical reaction of the cladding with water or steam must not exceed 0.01 times the
6 hypothetical amount that would be generated if all of the metal in the cladding cylinders
7 surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.
8 (iv) Coolable geometry. Calculated changes in core geometry must be such that the
9 core remains amenable to cooling.
10 (v) Long term cooling. After any calculated successful initial operation of the ECCS, the
11 calculated core temperature must be maintained at an acceptably low value and decay heat
12 must be removed for the extended period of time required by the long-lived radioactivity
13 remaining in the core.
14 (4) Acceptance criteria for LOCAs involving breaks larger than the TBS. The following
15 acceptance criteria must be used in determining the acceptability of ECCS cooling performance:
16 (i) Coolable geometry. Calculated changes in core geometry must be such that the core
17 remains amenable to cooling.
18 (ii) Long term cooling. After any calculated successful initial operation of the ECCS, the
19 calculated core temperature must be maintained at an acceptably low value and decay heat
20 must be removed for the extended period of time required by the long-lived radioactivity
21 remaining in the core.
22 (5) Imposition of restrictions. The Director of the Office of Nuclear Reactor Regulation or
23 the Office of New Reactors may impose restrictions on reactor operation if it is found that the
24 evaluations of ECCS cooling performance submitted are not consistent with paragraph (e) of
25 this section.
15 Revised: April 30, 2010
(f) Changes to facility, technical specifications, or procedures. 1 An applicant, permit
2 holder, or licensee or other entity who wishes to make changes enabled by this rule, to the
3 facility, facility design, or procedures or to the technical specifications shall perform a
4 risk-informed evaluation.
5 (1) The licensee may make changes, other than changes to the technical specifications,
6 without prior NRC approval if:
7 (i) The change is permitted under § 50.59 for holders of operating licenses, combined
8 licenses that do not reference a design certification, design approval, or manufacturing license
9 (per § 52.98(b)), or combined licenses that reference a design approval; permitted under
10 § 52.98(c) for holders of combined licenses that reference a design certification; or permitted
11 under § 52.98(d) for holders of combined licenses that reference a manufacturing license,
12 (ii) The risk informed evaluation process described in paragraph (c)(1)(iv) of this section
13 demonstrates that any increases in the estimated risk are minimal and the criteria in paragraph
14 (f)(3) of this section are met, and
15 (iii) The change does not invalidate the evaluation performed pursuant to paragraph
16 (c)(1)(i) of the applicability of the results in NUREG-1829 and NUREG-1903 to the licensee’s
17 facility.
18 (2) For implementing changes which are not permitted under paragraph (f)(1) of this
19 section, the permit holder, or licensee must submit an application for license amendment under
20 § 50.90. The application must contain:
21 (i) The information required under § 50.90;
22 (ii) Information from the risk-informed evaluation demonstrating that the total increases in
23 core damage frequency and large early release frequency are very small and the overall risk
24 remains small, and the criteria in paragraph (f)(3) of this section are met;
16 Revised: April 30, 2010
(iii) If previous changes 1 have been made under § 50.46a, information from the
2 risk-informed evaluation on the cumulative effect on risk of the proposed change and all
3 previous changes made under this section. If more than one plant change is combined;
4 including plant changes not enabled by this section, into a group for the purposes of evaluating
5 acceptable risk increases; the evaluation of each individual change shall be performed along
6 with the evaluation of combined changes; and
7 (iv) Information demonstrating that the criteria in paragraphs (e)(3) and (e)(4) of this
8 section are met.
9 (v) Information demonstrating that the proposed change will not increase the LOCA
10 frequency of the facility (including the frequency of seismically-induced LOCAs) by an amount
11 that would invalidate the applicability to the facility of the generic studies
12 (NUREG 1829,”Estimating Loss-of-Coolant Accident (LOCA) Frequencies through the
13 Elicitation Process”, March 2008 and NUREG-1903, “Seismic Considerations for the Transition
14 Break Size”, February 2008”) that support the technical basis for this section.
15 (3) All changes enabled by this rule must meet the following criteria:
16 (i) Adequate defense in depth is maintained;
17 (ii) Adequate safety margins are retained to account for uncertainties; and
18 (iii) Adequate performance-measurement programs are implemented to ensure the
19 risk-informed evaluation continues to reflect actual plant design and operation. These programs
20 shall be designed to detect degradation of the system, structure or component before plant
21 safety is compromised, provide feedback of information and timely corrective actions, and
22 monitor systems, structures or components at a level commensurate with their safety
23 significance, and
24
25 (iv) For applicants or licensees referencing a certified design, will not result in a
17 Revised: April 30, 2010
significant decrease in the level of safety otherwise pr 1 ovided by the certified design.
2 (4) Requirements for risk assessment - PRA. Whenever a PRA is used in the
3 risk-informed evaluation, the PRA must, with respect to the area of evaluation which is the
4 subject of the PRA:
5 (i) Address initiating events from sources both internal and external to the plant and for
6 all modes of operation, including low power and shutdown modes, that would affect the
7 regulatory decision in a substantial manner;
8 (ii) Reasonably represent the current configuration and operating practices at the plant;
9 (iii) Have sufficient technical adequacy (including consideration of uncertainty) and level
10 of detail to provide confidence that the total risk estimate and the change in total risk estimate
11 adequately reflect the plant and the effect of the proposed change on risk; and
12 (iv) Be determined, through peer review, to meet industry standards for PRA quality that
13 have been endorsed by the NRC.
14 (5) Requirements for risk assessment other than PRA. Whenever risk assessment
15 methods other than PRAs are used to develop quantitative or qualitative estimates of changes
16 to risk in the risk-informed evaluation, an integrated, systematic process must be used. All
17 aspects of the analyses must reasonably reflect the current plant configuration and operating
18 practices, and applicable plant and industry operating experience.
19 (g) Reporting.
20 (1) Licensees. (i) Each licensee shall estimate the effect of any change to or error in
21 evaluation models or analysis methods or in the application of such models or methods to
22 determine if the change or error is significant. For each change to or error discovered in an
23 ECCS evaluation model or analysis method or in the application of such a model that affects the
24
18 Revised: April 30, 2010
calculated results, the licensee shall report the nature of the change 1 or error and its estimated
2 effect on the limiting ECCS analysis to the Commission at least annually as specified in §§ 50.4
3 or 52.3. If the change or error is significant, the licensee shall provide this report within 30 days
4 and include with the report a proposed schedule for providing a reanalysis or taking other action
5 as may be needed to show compliance with § 50.46a requirements. This schedule may be
6 developed using an integrated scheduling system previously approved for the facility by the
7 NRC. For those facilities not using an NRC-approved integrated scheduling system, a schedule
8 will be established by the NRC staff within 60 days of receipt of the proposed schedule. Any
9 change or error correction that results in a calculated ECCS performance that does not conform
10 to the criteria set forth in paragraphs (e)(3) or (e)(4) of this section is a reportable event as
11 described in §§ 50.55(e), 50.72 and 50.73. The licensee shall propose immediate steps to
12 demonstrate compliance or bring plant design or operation into compliance with § 50.46a
13 requirements. For the purpose of this paragraph, a significant change or error is:
14 (A) For LOCAs involving pipe breaks at or below the TBS, one which results either in a
15 calculated peak fuel cladding temperature different by more than 50°F from the temperature
16 calculated for the limiting transient using the last acceptable model, or is a cumulation of
17 changes and errors such that the sum of the absolute magnitudes of the respective temperature
18 changes is greater than 50°F; or
19 (B) For LOCAs involving pipe breaks larger than the TBS, one which results in a
20 significant reduction in the capability to meet the requirements of paragraph (e)(4) of this
21 section.
22 (ii) As part of the PRA maintenance and upgrading under paragraph (d)(4) of this
23 section, the licensee shall report to the NRC if the re-evaluation results in exceeding the
24 acceptance criteria in paragraph (f) of this section, as applicable. The report must be filed with
25
19 Revised: April 30, 2010
the NRC no more than 60 days after completing the PRA r 1 e-evaluation. The report must
2 describe and explain the changes in the PRA modeling, plant design, or plant operation that led
3 to the increase(s) in risk, and must include a description of and implementation schedule for any
4 corrective actions required under paragraph (d)(4) of this section.
5 (iii) Every 24 months, the licensee shall submit, as specified in §§ 50.4 or 52.3, a short
6 description of each change involving minimal changes in risk made under paragraph (f)(1) of
7 this section after the last report and a brief summary of the basis for the licensee’s
8 determination pursuant to § 50.46a(f)(2)(vi) that the change does not invalidate the applicability
9 evaluation made under § 50.46a(c)(1)(i).
10 (2) Design certifications; applicants for and holders of design approvals. Each design
11 certification applicant and each applicant for and holder of a design approval shall report to the
12 NRC any significant error in evaluation models or analysis methods or in the application of such
13 models or methods. A design certification applicant’s duty to report under this paragraph
14 continues until the later of either the termination or expiration of the design certification; or the
15 termination of the last license directly or indirectly referencing the design certification. For the
16 purpose of this paragraph, a significant change or error is:
17 (i) For LOCAs involving pipe breaks at or below the TBS, one which results either in a
18 calculated peak fuel cladding temperature different by more than 50 °F from the temperature
19 calculated for the limiting transient using the last acceptable model, or is a cumulation of
20 changes and errors such that the sum of the absolute magnitudes of the respective temperature
21 changes is greater than 50 °F; or
22 (ii) For LOCAs involving pipe breaks larger than the TBS, one which results in a
23 significant reduction in the capability to meet the requirements of paragraph (e)(4) of this
24 section.
20 Revised: April 30, 2010
(h) Documentation. Following implementation of 1 the § 50.46a requirements, each entity
2 subject to this section shall maintain records sufficient to demonstrate compliance with the
3 requirements in this section in accordance with § 50.71.
4 (i) through (l) - [RESERVED]
5 (m) Changes to TBS.
6 (1) Operating licenses under Part 50, combined licenses under Part 52, and
7 manufacturing licenses. If the NRC increases the TBS specified in this section, each licensee
8 subject to this section (other than a licensee referencing a design certification rule complying
9 with the requirements of this section) shall re-perform the evaluations required by paragraphs
10 (e)(1) and (e)(2) of this section and reconfirm compliance with the acceptance criteria in
11 paragraphs (e)(3) and (e)(4) of this section. If the licensee cannot demonstrate compliance with
12 the acceptance criteria, then the licensee shall change its facility, technical specifications or
13 procedures so that the acceptance criteria are met. The evaluation required by this paragraph,
14 and any necessary changes to the facility, technical specifications or procedures as the result of
15 this evaluation, are not to be deemed to be backfitting under any provision of this chapter or a
16 violation of any finality provision in Part 52.
17 (2) Design certifications and referencing combined licenses. Changes to a TBS for a
18 design certification must be accomplished by rulemaking, in accordance with 10 CFR 52.63(a).
19 Holders of combined licenses referencing a design certification rule shall re perform the
20 evaluations required by paragraphs (e)(1) and (e)(2) of this section and reconfirm compliance
21 with the acceptance criteria in paragraphs (e)(3) and (e)(4) of this section. If the licensee cannot
22 demonstrate compliance with the acceptance criteria, then the licensee shall change its facility,
23 technical specifications or procedures so that the acceptance criteria are met. These actions are
24 deemed to be in conformance with applicable finality provisions in Part 52.
25
21 Revised: April 30, 2010
5. In ' 50.109, paragraph (b) 1 is revised to read as follows:
2 ' 50.109 Backfitting.
3
4 * * * * *
5
6 (b) Paragraph (a)(3) of this section shall not apply to:
7 (1) Backfits imposed prior to October 21, 1985; and
8 (2) Any changes made to the TBS specified in ' 50.46a or as otherwise applied to a
9 licensee.
10 * * * * *
11 6. In Appendix A to 10 CFR Part 50, under the heading, ACRITERIA,@ Criterion 17, 35,
12 38, 41, 44, and 50 are revised to read as follows:
13 APPENDIX A TO PART 50 -GENERAL DESIGN CRITERIA FOR NUCLEAR POWER
14 PLANTS
15
16 * * * * *
17 CRITERIA
18 * * * * *
19
20 Criterion 17--Electrical power systems. An on-site electric power system and an offsite
21 electric power system shall be provided to permit functioning of structures, systems, and
22 components important to safety. The safety function for each system (assuming the other
23 system is not functioning) shall be to provide sufficient capacity and capability to assure that (1)
24 specified acceptable fuel design limits and design conditions of the reactor coolant pressure
22 Revised: April 30, 2010
boundary are not exceeded as a result of 1 anticipated operational occurrences and (2) the core
2 is cooled and containment integrity and other vital functions are maintained in the event of
3 postulated accidents.
4 The onsite electric power supplies, including the batteries, and the onsite electrical
5 distribution system, shall have sufficient independence, redundancy, and testability to perform
6 their safety functions assuming a single failure, except for loss of coolant accidents involving
7 pipe breaks larger than the transition break size under ' 50.46a, where a single failure of the
8 onsite power supplies and electrical distribution system need not be assumed for plants under
9 ' 50.46a. For those pipe breaks only, neither a single failure nor the unavailability of offsite
10 power need be assumed.
11 Electric power from the transmission network to the onsite electric distribution system
12 shall be supplied by two physically independent circuits (not necessarily on separate rights of
13 way) designed and located so as to minimize to the extent practical the likelihood of their
14 simultaneous failure under operating and postulated accident conditions. A switchyard common
15 to both circuits is acceptable. Each of these circuits shall be designed to be available in
16 sufficient time following a loss of all onsite alternating current power supplies and the other
17 offsite electric power circuit, to assure that specified acceptable fuel design limits and design
18 conditions of the reactor coolant pressure boundary are not exceeded. One of these circuits
19 shall be designed to be available within a few seconds following a LOCA to assure that core
20 cooling, containment integrity, and other vital safety functions are maintained.
21 Provisions shall be included to minimize the probability of losing electric power from any
22 of the remaining supplies as a result of, or coincident with, the loss of power generated by the
23 nuclear power unit, the loss of power from the transmission network, or the loss of power from
24 the onsite electric power supplies.
25
23 Revised: April 30, 2010
1 * * * * *
2
3 Criterion 35--Emergency core cooling. A system to provide abundant emergency core
4 cooling shall be provided. The system safety function shall be to transfer heat from the reactor
5 core following any loss of reactor coolant at a rate such that (1) fuel and clad damage that could
6 interfere with continued effective core cooling is prevented and (2) clad metal-water reaction is
7 limited to negligible amounts.
8 Suitable redundancy in components and features, and suitable interconnections, leak
9 detection, isolation, and containment capabilities shall be provided to assure that for onsite
10 electric power system operation (assuming offsite power is not available) and for offsite electric
11 power system operation (assuming onsite power is not available) the system safety function can
12 be accomplished, assuming a single failure, except for loss of coolant accidents involving pipe
13 breaks larger than the transition break size under ' 50.46a. For those pipe breaks only, neither
14 a single failure nor the unavailability of offsite power need be assumed.
15
16 * * * * *
17
18 Criterion 38--Containment heat removal. A system to remove heat from the reactor
19 containment shall be provided. The system safety function shall be to reduce rapidly, consistent
20 with the functioning of other associated systems, the containment pressure and temperature
21 following any LOCA and maintain them at acceptably low levels.
22 Suitable redundancy in components and features, and suitable interconnections, leak
23 detection, isolation, and containment capabilities shall be provided to assure that for onsite
24 electric power system operation (assuming offsite power is not available) and for offsite electric
24 Revised: April 30, 2010
power system operation (assuming onsite 1 power is not available) the system safety function can
2 be accomplished, assuming a single failure, except for analysis of loss of coolant accidents
3 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
4 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
5
6 * * * * *
7
8 Criterion 41--Containment atmosphere cleanup. Systems to control fission products,
9 hydrogen, oxygen, and other substances which may be released into the reactor containment
10 shall be provided as necessary to reduce, consistent with the functioning of other associated
11 systems, the concentration and quality of fission products released to the environment following
12 postulated accidents, and to control the concentration of hydrogen or oxygen and other
13 substances in the containment atmosphere following postulated accidents to assure that
14 containment integrity is maintained.
15 Each system shall have suitable redundancy in components and features, and suitable
16 interconnections, leak detection, isolation, and containment capabilities to assure that for onsite
17 electric power system operation (assuming offsite power is not available) and for offsite electric
18 power system operation (assuming onsite power is not available) its safety function can be
19 accomplished, assuming a single failure, except for analysis of loss of coolant accidents
20 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
21 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
22
23 * * * * *
24
25 Revised: April 30, 2010
Criterion 44--Cooling water. A system to transfer heat 1 from structures, systems, and
2 components important to safety, to an ultimate heat sink shall be provided. The system safety
3 function shall be to transfer the combined heat load of these structures, systems, and
4 components under normal operating and accident conditions.
5 Suitable redundancy in components and features, and suitable interconnections, leak
6 detection, and isolation capabilities shall be provided to assure that for onsite electric power
7 system operation (assuming offsite power is not available) and for offsite electric power system
8 operation (assuming onsite power is not available) the system safety function can be
9 accomplished, assuming a single failure, except for analysis of loss of coolant accidents
10 involving pipe breaks larger than the transition break size under ' 50.46a. For those pipe
11 breaks only, neither a single failure nor the unavailability of offsite power need be assumed.
12
13 * * * * *
14
15 Criterion 50--Containment design basis. The reactor containment structure, including
16 access openings, penetrations, and the containment heat removal system shall be designed so
17 that the containment structure and its internal compartments can accommodate, without
18 exceeding the design leakage rate and with sufficient margin, the calculated pressure and
19 temperature conditions resulting from any loss-of-coolant accident. This margin shall reflect
20 consideration of (1) the effects of potential energy sources which have not been included in the
21 determination of the peak conditions, such as energy in steam generators and as required by
22 ' 50.44 energy from metal-water and other chemical reactions that may result from degradation
23 but not total failure of emergency core cooling functioning, (2) the limited experience and
24 experimental data available for defining accident phenomena and containment responses, and
25
26 Revised: April 30, 2010
(3) the conservatism 1 of the calculational model and input parameters.
2 For licensees voluntarily choosing to comply with ' 50.46a, the structural and leak tight
3 integrity of the reactor containment structure, including access openings, penetrations, and its
4 internal compartments, shall be maintained for realistically calculated pressure and temperature
5 conditions resulting from any loss of coolant accident larger than the transition break size.
6
7 * * * * *
8
9 PART 52 - LICENSES, CERTIFICATIONS AND APPROVALS FOR NUCLEAR POWER
10 PLANTS
11 7. The authority citation for part 52 continues to read as follows:
12 AUTHORITY: Secs. 103, 104, 161, 182, 183, 185, 186, 189, 68 Stat. 936, 948, 953, 954,
13 955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2133, 2201, 2232, 2233,
14 2235, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended
15 (42U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note); Energy Policy
16 Act of 2005, Pub. L. No. 109-58, 119 Stat. 594 (2005), secs. 147 and 149 of the Atomic Energy
17 Act.
18 8. In § 52.47, paragraph (a)(4) is revised to read as follows:
19 § 52.47 Contents of applications; technical information
20
21 (a) * * *
22
23 (4) An analysis and evaluation of the design and performance of structures, systems,
24 and components with the objective of assessing the risk to public health and safety resulting
27 Revised: April 30, 2010
from operation of the facility and including determination of the 1 margins of safety during normal
2 operations and transient conditions anticipated during the life of the facility, and the adequacy of
3 structures, systems, and components provided for the prevention of accidents and the mitigation
4 of the consequences of accidents.
5 (i) Analysis and evaluation of emergency core cooling system (ECCS) cooling
6 performance and the need for high-point vents following postulated loss-of-coolant accidents
7 may be performed under the requirements of either § 50.46 or § 50.46a and § 50.46b of this
8 chapter for designs certified after [EFFECTIVE DATE OF RULE] and demonstrated under
9 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
10 DATE OF RULE], or
11 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
12 vents following postulated loss-of-coolant accidents must be performed under the requirements
13 of §§ 50.46 and 50.46b of this chapter for designs that are not demonstrated under
14 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
15 DATE OF RULE].
16
17 * * *
18
19 § 52.54 Issuance of standard design certification.
20
21 * * *
22
23 (b) The design certification rule must specify the site parameters, design
24 characteristics, and any additional requirements and restrictions of the design certification rule.
28 Revised: April 30, 2010
A design certification rule which was reviewed and appr 1 oved as meeting the requirements of
2 10 CFR 50.46a must specify the criteria governing departures that a referencing combined
3 license must meet. The criteria must ensure that the safety bases for the NRC’s approval of the
4 certified design’s compliance with § 50.46a (including applicability of the TBS) continue to apply
5 despite the departure.
6
7 * * * * *
8
9 9. In § 52.79, paragraph (a)(5) is revised to read as follows:
10 § 52.79 Contents of applications; technical information in final safety analysis report.
11
12 (a) * * *
13
14 (5) An analysis and evaluation of the design and performance of structures, systems,
15 and components with the objective of assessing the risk to public health and safety resulting
16 from operation of the facility and including determination of the margins of safety during normal
17 operations and transient conditions anticipated during the life of the facility, and the adequacy of
18 structures, systems, and components provided for the prevention of accidents and the mitigation
19 of the consequences of accidents.
20 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
21 vents following postulated loss-of-coolant accidents must be performed under the requirements
22 of either § 50.46 or § 50.46a and § 50.46b of this chapter for facilities licensed after
23 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) of this chapter to be
24 similar to reactor designs licensed before [EFFECTIVE DATE OF RULE], or
29 Revised: April 30, 2010
(ii) Analysis and evaluation of 1 ECCS cooling performance and the need for high-point
2 vents following postulated loss-of-coolant accidents must be performed under the requirements
3 of §§ 50.46 and 50.46b of this chapter for facilities licensed after [EFFECTIVE DATE OF RULE]
4 and not demonstrated under § 50.46a(c)(2) of this chapter to be similar to reactor designs
5 licensed before [EFFECTIVE DATE OF RULE].
6
7 * * * * *
8
9 10. In § 52.137, paragraph (a)(4) is revised to read as follows:
10 § 52.137 Contents of applications; technical information.
11
12 (a) * * *
13
14 (4) An analysis and evaluation of the design and performance of SSCs with the objective
15 of assessing the risk to public health and safety resulting from operation of the facility and
16 including determination of the margins of safety during normal operations and transient
17 conditions anticipated during the life of the facility, and the adequacy of SSCs provided for the
18 prevention of accidents and the mitigation of the consequences of accidents.
19 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
20 vents following postulated loss-of-coolant accidents must be performed under the requirements
21 of either § 50.46 or § 50.46a and § 50.46b of this chapter for designs approved after
22 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) of this chapter to be
23 similar to reactor designs licensed before [EFFECTIVE DATE OF RULE], or
24 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
30 Revised: April 30, 2010
vents following postulated loss-of-coolant accidents must be 1 performed under the requirements
2 of §§ 50.46 and 50.46b of this chapter for designs that are not demonstrated under
3 § 50.46a(c)(2) of this chapter to be similar to reactor designs licensed before [EFFECTIVE
4 DATE OF RULE].
5
6 * * * * *
7
8 11. In § 52.157, paragraph (f)(1) is revised to read as follows:
9 § 52.157 Contents of applications; technical information in final safety analysis report.
10
11 (f) * * *
12
13 (1) An analysis and evaluation of the design and performance of structures, systems,
14 and components with the objective of assessing the risk to public health and safety resulting
15 from operation of the facility and including determination of the margins of safety during normal
16 operations and transient conditions anticipated during the life of the facility, and the adequacy of
17 structures, systems, and components provided for the prevention of accidents and the mitigation
18 of the consequences of accidents.
19 (i) Analysis and evaluation of ECCS cooling performance and the need for high-point
20 vents following postulated loss-of-coolant accidents must be performed under the requirements
21 of either § 50.46 or § 50.46a and § 50.46b of this chapter for facilities licensed after
22 [EFFECTIVE DATE OF RULE] and demonstrated under § 50.46a(c)(2) to be similar to reactor
23 designs licensed before [EFFECTIVE DATE OF RULE], or
24 (ii) Analysis and evaluation of ECCS cooling performance and the need for high-point
31 Revised: April 30, 2010
vents following postulated loss-of-coolant accidents must be 1 performed under the requirements
2 of §§ 50.46 and 50.46b of this chapter for facilities licensed after [EFFECTIVE DATE OF RULE]
3 and not demonstrated under § 50.46a(c)(2) of this chapter to be similar to reactor designs
4 licensed before [EFFECTIVE DATE OF RULE].
5
6 * * * * *
7
8 Dated at Rockville, Maryland, this day of , 2010.
9 For the Nuclear Regulatory Commission.
10
11
12
13
14 R. W. Borchardt,
15 Executive Director
16 for Operations
17
18