ML14035A554

NRR E-mail Capture - Reminder: PRB Pre-Meeting Phoncon on Tuesday, 2/4

ML14035A554
Person / Time
Site:	Columbia
Issue date:	02/04/2014
From:	Clay Johnson - No Known Affiliation
To:	Lyon F Division of Operating Reactor Licensing
References
G20130776, MF3031
Download: ML14035A554 (24)

Text

NRR-PMDAPEm Resource From: johnsonc20@gmail.com on behalf of Charles K. Johnson [chuck@oregonpsr.org]

Sent: Tuesday, February 04, 2014 12:16 PM To: Lyon, Fred Cc: Banic, Merrilee; Nancy Matela

Subject:

Re: REMINDER: PRB Pre-Meeting Phoncon on Tuesday, 2/4 Attachments: CGS Hz Question 30Jan2014-1.docx; Resp Letter to ENW S - 10Jan14 TLT-1.docx; 20111003-ucs-brief-bwr-scram-problem.pdf; 20081113-cook-ler-turbine-failure.pdf; Dave Lochbaum 1-29-14 email - portion - regarding implications of DC Cook plant fire accident.doc Hi Fred & Merrilee, Here are (attached, in order):

1) Terry Tolan's question to the authors (Yong Li or Cliff Munson) of the internal determination NRC memo responding to his original analysis of the seismic issues related to CGS' safety in a worst case earthquake;

2) Terry Tolan's response letter to Charles Johnson regarding the NRC's internal determination memo to reject O&WPSR's call for immediate shutdown of the CGS pending earthquake assessment;

3) David Lochbaum's issue brief for the Union of Concerned Scientists entitled "Boiling Water Reactor Shut Down System Problem" regarding GE's acknowledge problem that the control blades may not insert properly into control channels during an earthquake - preventing shut down;

4) NRC Licensee Event Report 315/2008-006-00 "Manual Reactor Trip Due to Main Turbine High Vibration" at the Donald C. Cook Nuclear Plant Unit 1, and,

5) Dave Lochbaum 1-29-14 email to Charles Johnson (portion) regarding implications of DC Cook plant fire event when considering the consequences of a SSE-magnitude quake on fire protection systems.

I look forward to speaking with you and your colleagues momentarily.

All the best, Chuck Johnson On Tue, Feb 4, 2014 at 7:17 AM, Lyon, Fred <Fred.Lyon@nrc.gov> wrote:

P.S. Either before or after the call, please email the question, memo, fact sheet, et al., to me. We put everything into ADAMS and make it available to the public (except for allegations, any personally identifiable information, e.g., social security numbers, and such). Also, if the Alliance wants to join you, then they also need to send us a letter, or email me. It doesnt count in the 2.206 process if its not on paper (at the very least, electronically). Thanks, Fred From: Charles K. Johnson [1]

Sent: Monday, February 03, 2014 5:53 PM To: Lyon, Fred

Subject:

Re: REMINDER: PRB Pre-Meeting Phoncon on Tuesday, 2/4 Hi Fred, 1

Tomorrow I will be on the call as will Nancy Matela from the Alliance for Democracy here in Portland. The Alliance for Democracy is joining us in the petition.

We will submit a question from geologist Terry Tolan, a new memo from Mr. Tolan, a fact sheet from David Lochbaum of UCS on channel control blade interference, and a document relating to the fire in the non-nuclear portion of the Cook nuclear power plant in Illinois that we believe may have some bearing on earthquake related hazards at the CGS.

We reserve the right to use the entire 50 minutes, but expect the time will be considerably less than that.

All the best, Chuck Johnson On Sat, Feb 1, 2014 at 3:44 PM, Lyon, Fred <Fred.Lyon@nrc.gov> wrote:

No problem. Thanks, Chuck, Fred From: Charles K. Johnson [johnsonc20@gmail.com]

Sent: Friday, January 31, 2014 6:53 PM To: Lyon, Fred

Subject:

Re: REMINDER: PRB Pre-Meeting Phoncon on Tuesday, 2/4 Hi Fred, Sorry for being incommunicado. I am still working on the total number of people and what we will be presenting on Tuesday. I will let you know the names and identifiers on Monday afternoon. It will probably be a small number of people and will not likely use all of our 50 minutes.

All the best, Chuck On Fri, Jan 31, 2014 at 5:31 AM, Lyon, Fred <Fred.Lyon@nrc.gov<mailto:Fred.Lyon@nrc.gov>> wrote:

Chuck, please let me know the folks you expect on the call on your end. Remember, you have about 50 minutes to present your information, and you can have anyone you want speak, in whatever order you desire. You talk, we listen. Its a recorded call and a transcript will be made and put into our ADAMS public document system.

If youve any questions, please give me a call or email. Thanks, Fred Bridge number (800) 772-3842<tel:%28800%29%20772-3842>, passcode 2206. The call is scheduled for 1-2 PM EST.

The purpose of the telecon is for the petitioner to provide any relevant additional information and support for the request, letter from OWPSR to NRC dated 10/31/13

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2

Charles K. Johnson 5031 SE Haig St.

Portland, OR 97206 (503) 777-2794 johnsonc20@gmail.com<mailto:johnsonc20@gmail.com>

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Charles K. Johnson 5031 SE Haig St.

Portland, OR 97206 (503) 777-2794 johnsonc20@gmail.com

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Charles K. Johnson Director, Joint Task Force on Nuclear Power Oregon and Washington Physicians for Social Responsibility 812 SW Washington Street, Suite 1050 Portland, OR 97205 (503) 777-2794 cell chuck@oregonpsr.org 3

Hearing Identifier: NRR_PMDA Email Number: 1057 Mail Envelope Properties (CAHWtKjfDCiRzrs_P0cPhMF1Pu7CYPFNjHXVHh14-0+in+S9wDA)

Subject:

Re: REMINDER: PRB Pre-Meeting Phoncon on Tuesday, 2/4 Sent Date: 2/4/2014 12:16:07 PM Received Date: 2/4/2014 12:16:54 PM From: johnsonc20@gmail.com Created By: johnsonc20@gmail.com Recipients:

"Banic, Merrilee" <Merrilee.Banic@nrc.gov>

Tracking Status: None "Nancy Matela" <nancy.matela@gmail.com>

Tracking Status: None "Lyon, Fred" <Fred.Lyon@nrc.gov>

Tracking Status: None Post Office: mail.gmail.com Files Size Date & Time MESSAGE 5207 2/4/2014 12:16:54 PM CGS Hz Question 30Jan2014-1.docx 105152 Resp Letter to ENW S - 10Jan14 TLT-1.docx 25902 20111003-ucs-brief-bwr-scram-problem.pdf 487809 20081113-cook-ler-turbine-failure.pdf 317918 Dave Lochbaum 1-29-14 email - portion - regarding implications of DC Cook plant fire accident.doc 28736 Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received:

Question:

Is it the case that the maximum vibratory ground motion (SSE) for the Columbia Generating Station is 0.25g to 0.60g in the 2 to 10 Hertz (Hz) range on Figure 1 as stated in the attached letter? If so, can you explain the statement on page 2 of the letter (highlighted in yellow) that cites 20 Hz? Should it state 2Hz and greater?

Columbia Generating Station Seismic Hazard Considerations Determination of Immediate Safety Concerns G20130776

Background:

• Seismic designs at US nuclear power plants are developed in terms of seismic ground motion spectra, which are called the Safe Shutdown Earthquake ground motion response spectra (SSE).

• Each nuclear power plant is designed to a ground motion level that is appropriate for the geology and tectonics in the region surrounding the plant location.

• Currently operating nuclear power plants developed their SSEs based on a deterministic or scenario earthquake that accounts for the largest earthquake expected in the area around the plant.

• The SSE for operating nuclear power plants in the U.S. is based on that earthquake which produces the maximum vibratory ground motion for which key structures, systems, and components are designed to remain functional.

• Due to code requirements and design standards, nuclear power plants are designed and built to have seismic margins generally well beyond the SSE level.

Evaluation:

• The seismic design for the Columbia Generating Station (CGS) plant is represented by the SSE ground motion response spectrum, as shown in Figure 1 below (solid curve).

o The CGS SSE spectrum is anchored at an acceleration level of 0.25 g, but is much higher (up to 0.6 g) over the important frequency range of 2 to 10 Hz where plant structures and systems are most sensitive to earthquake ground motions.

• Since the CGS operating license was issued in 1984, the licensee has reevaluated the seismic hazards for the plant as part of the Individual Plant Examination of External Events (IPEEE) program in the early-to mid-1990s.

o Under the IPEEE program, the licensee conducted a full probabilistic seismic hazard analysis for the region around the CGS plant, including an evaluation of earthquake activity in the Columbia Basin (including the Yakima fold belt) and the Cascadia subduction zone.

o The licensee evaluated the impact on the CGS plant from potential seismic ground motions from several regional seismic sources over a wide range of hazard levels (the 1/10,000 mean hazard levels are shown as circles in Figure 1 below).

o The licensee evaluated the seismic capacity or ruggedness of the CGS plant and determined that the risk of core damage from a seismic event is very low (2x10-5

per year).

• The Department of Energy (DOE) evaluated the seismic hazards for the seismic design of the Waste Treatment and Immobilization Plant (WTP) at the Hanford site in 1995, 2005, and 2007.

o The most recent WTP seismic design spectra are shown below as the dashed (2005) and dotted (2007) curves.

o The most recent seismic hazard evaluation for the design of the WTP in 2007 is very similar to the CGS seismic design or SSE.

o In its letter to the Chairman dated October 31, 2013, the Oregon and Washington Physicians for Social Responsibility (OWPSR) mistakenly compares the 3 to 5 Hz spectral acceleration level of 0.8g for the WTP 2005 seismic design with the CGS SSE 20 Hz and greater spectral acceleration value of 0.25g.

• Under Fukushima Near-Term Task Force Recommendation 2.1 (NTTF R2.1), the licensee (Energy Northwest) is currently reevaluating (along with DOE) the seismic hazards for the region surrounding the Hanford site using the latest data, models and methods, consistent with current NRC regulatory guidance.

o NTTF R2.1 specifies that the licensee will need to evaluate all of the potential seismic sources (including the Umtanum and Yakima Ridge faults) in the site region.

o All of the issues raised in the letter from OWPSR are known and are being evaluated as part of the seismic hazard reevaluation being conducted by DOE and Energy Northwest.

o If the reevaluated hazard is greater than the CGS plant seismic design or SSE, the licensee will perform a complete seismic risk evaluation for the plant as well as important interim actions while the risk evaluation is ongoing.

o The reevaluated hazard evaluation is due to the NRC in March 2015.

==

Conclusion:==

Based on the information discussed above, the NRC staff concludes that there is no immediate safety concern at CGS; however, the NRC will review the seismic hazard and risk evaluations conducted by Energy Northwest for potential regulatory action as part of its evaluation for NTTF R2.1.

Figure 1: SSE GROUND MOTION RESPONSE SPECTRUM Principal Contributors: Yong Li, NRR/DE/EMCB Cliff Munson, NRO/DSEA Date: December 18, 2013

January 7, 2014 Mr. Charles Johnson Washington Physicians for Social Responsibility RE: Some comments on the Columbia Generating Station Seismic Hazard Considerations Determination of Immediate Safety Concerns letter (G20130776)

Dear Mr. Johnson:

I have had an opportunity to read the response letter cited above. I would first like to respond to the Background information set out by the NRC staff.

The NRC staff states that each nuclear power plant is designed to a ground motion level that is appropriate for the geology and tectonics in the region surrounding the plant location. In my two letter reports (referenced at the end of the letter), it has been clearly shown that the CGS was designed at a time when much of the geology and tectonics of this region was not well understood. For example:

- the CGS was designed based on the flawed decision that the epicenter of the largest historical earthquake in this region was 180 miles away as compared to recent study that places the epicenter approximately 99 miles away from the CGS site. This is a significant change that needs to be assessed.

- When the CGS was designed, only 6 faults were considered within the Yakima Fold and Thrust Belt as part of the seismic risk assessment. Subsequent studies have more than doubled this number of Yakima Fold and Thrust Belt faults.

- The design of the CGS was based on a thin-skin (uncoupled) fault model for Yakima Fold and Thrust Belt. Published work by the U.S.

Geologic Survey (Blakely et al., 2011) shows that the Yakima Fold and Thrust Belt is best characterized by a thick-skin (coupled) fault model which means that the faulting extends into the basement Page 1 of 4

rock below basalt layers and can give rise to larger magnitude earthquakes and subsequently higher vibratory ground motion at the CGS site.

- When designed the CGS seismic risk assessment was based on Yakima Fold and Thrust Belt faults that had relatively short lengths.

Based on the published U.S. Geological Survey work (Blakely et al.,

2011) it appears that these faults are substantially longer and likely capable of producing much larger magnitude earthquakes than previously believed.

- When the CGS was designed, it was thought that movement on the Yakima Fold and Thrust Belt faults was old due to the lack of evidence of young offsets. This assumption is also questionable based on the published work by Blakely et al. (2011).

- When the CGS was designed, it was not known that there was an active fault just 2.3 miles away from the reactor.

Thus I question the NRC staff assumption that the design of CGS is appropriate for the geology and tectonics in the region given all of the fundamental revisions in our knowledge that have occurred since the last time the CGS seismic risk was assessed in the early 1990s. These advances in understanding the structural geology of the Yakima Fold and Thrust Belt fundamental change the basic assumptions previously used to assess seismic hazards and risk at the CGS site.

Concerning the Evaluation, I note that they state on Page 2 (first bullet, third item) that OWPSR mistakenly compared the 3 to 5 Hz spectral acceleration level of 0.8g for the WTP 2005 seismic design with the CGS SSB 20 Hz and greater spectral acceleration value of 0.25g. Indeed I should have compared the CGS spectral acceleration to that of the revised WTP spectral response developed by Young (2007) which is 0.6 g. However, it still doesn't change the fact that the maximum vibratory ground motion for this area has been dramatically increased based on the WTP studies.

As shown in their Figure 1 (Page 3 of the letter), the revised WTP response spectrum is very similar to that developed for the CGS. This is interesting given that Energy Northwest previously implied (letter to the NRC dated 17 September 2010; response to 3a, second paragraph) that differences in Page 2 of 4

site factors (e.g., distance from active faults, physical soil properties and thicknesses, amplification/deamplification, etc.) between the CSG and WTP sites does not allow one to apply any of the recent WTP seismic ground motion findings (Rohay and Reidel, 2005, Rohay and Brouns, 2007; Young, 2007) to the CGS site. It is curious that if the site factors are as different between the CSG and WTP as Energy Northwest claimed, why are the vibratory ground motion response spectrums so very similar?

As previously noted and discussed, Energy Northwest needs to develop a CGS site-specific model for ground motion response spectrum based on borehole vertical seismic profile data from the ground surface to the top of the Columbia River basalt. They then need to integrate this data with the WTP shear wave velocity data for the Columbia River basalt/Ellensburg Formation interbeds. Energy Northwest also needs to reevaluate the maximum credible earthquakes and overall seismic hazards in light of the recently published U.S. Geological Survey work (Blakely et al., 2011). They would have to incorporate a coupled fault model, extended active fault lengths, and reevaluate the earthquake magnitude/frequency, etc.) before the CGS site-specific subsurface velocity data could be used to help constrain estimates of vibratory ground motion from various earthquake scenarios.

In the last bullet paragraph (Page 2) it is stated that Energy Northwest, along with the U.S. Department of Energy, is reevaluating the seismic hazards for the region surrounding the Hanford site and is using the latest data, models and methods. They also indicated that they are evaluating the issues we have previously raised as part of this new seismic hazard reevaluation and that this report is due to the NRC in March 2015. I look forward to seeing this report.

In summary, since both the U.S. Department of Energys (Youngs, 2007; Rohay and Brouns, 2007; Rohay and Reidel, 2005) and Energy Northwests seismic hazard analyses rely on the flawed and outmoded seismic assessment model developed by Geomatrix (1996), one needs to question the basic adequacy of the existing CGS seismic hazards analysis in light of the new and recent data and findings presented by the U.S. Geological Survey.

Sincerely, Terry L. Tolan, LEG Page 3 of 4

Letter Reports Cited:

Earthquake risk factors at the Columbia Generating Station dated October 31, 2013 by Terry L. Tolan, LEG.

Evaluation of Energy Northwest Response (letter dated 17 September 2010) to Nuclear Regulatory Commission Request 3a (letter dated 13 July 2010) for Additional Information on Seismic Hazards for the Review of the Columbia Generating Station License Renewal Applicationdated October 31, 2013 by Terry L. Tolan, LEG.

References Cited:

Blakely,, R.J., Sherrod, B.L., Weaver, C.S., Wells, R.E., Rohay, A.C., Barnett, E.A., and Knepprath, N.E., 2011, Connecting the Yakima fold and thrust belt to active faults in the Puget Lowland, Washington: Journal of Geophysical Research, v. 116, B07105, 33 p., doi:10.1029/2010Jb008091.

Geomatrix, 1996, Probabilistic seismic hazard analysis, DOE Hanford Site, Washington: prepared by Geomatrix Consultants, Inc., for Westinghouse Hanford Company, Richland, Washington, WHC-SD-W236-TI-002, Rev. 1a.

Rohay, A.C., and Reidel, S.P., 2005, Site-specific seismic response model for the Waste Treatment Plant, Hanford, Washington: Battelle Pacific Northwest National Laboratory, Richland, Washington, PNNL-15089, 160 p.

Rohay, A.C., and Brouns, T.M., 2007, Site-specific velocity and density model for the Waste Treatment Plant, Hanford, Washington: Battelle Pacific Northwest National Laboratory, Richland, Washington, PNNL-16652, 76 p.

Youngs, R.R, 2007, Updated site response analysis for the Waste Treatment Plant, DOE Hanford Site, Washington: prepared by Geomatrix Consultants, Inc., for the Battelle Pacific Northwest National Laboratory, Richland, Washington, PNNL-16653 (GMX-9995.002-001), 47 p.

Page 4 of 4

Boiling Water Reactor Shut Down System Problem GE Hitachi informed the Nuclear Regulatory Commission (NRC) about a safety problem related to the reactor shut down system at its boiling water reactors (BWRs) via a September 27, 2011 update to NRC Event Report No. 46230 dated September 3, 2010:

GE Hitachi (GEH) has determined that the scram capability of the control rod drive mechanism in BWR/2-5 plants may not be sufficient to ensure the control rod will fully insert in a cell with channel-control rod friction at or below the friction limits specified in MFN 08-420 with a concurrent Safe Shutdown Earthquake (SSE). The plant condition for which incomplete control rod insertion might occur is when the reactor is below normal operating pressure (<900 psig) and a scram occurs concurrent with the SSE, for Mark I containment plants, and for the SSE with concurrent Loss-of-Coolant Accident (LOCA) and Safety Relief Valve (SRV) events for Mark II containment plants. In this scenario a Substantial Safety Hazard results because the affected control rods might not fully insert to perform the required safety function.

GEH has determined that when channel-control blade interference is present at reduced reactor pressure and at friction levels considered acceptable in MFN 08-420, a simultaneously occurring Safe Shutdown Earthquake (SSE) may result in control rod friction that inhibits the full insertion of the affected control rods during a reactor scram from these conditions. This scenario was not explicitly considered in MFN 08-420.

This issue brief provides background information about control rods at BWRs and this specific safety problem.

The reactor core of a BWR consists of fuel pellets stacked within hollow metal rods that are formed into fuel assemblies. Each fuel assembly is housed within a hollow metal box called a channel. The BWR core has over 100 fuel cells, each consisting of four fuel assemblies with one control rod in the middle.

The BWR core is powered by the fissioning, or splitting, of atoms in the fuel pellets. Energy, and sub-atomic particles called neutrons, are released when certain atoms split. The energy boils water flowing past the fuel assemblies, hence the name boiling water reactor. The neutrons interact with other atoms, causing them to become unstable and split to sustain the nuclear chain reaction.

Control rods function as both the gas pedal and the brake pedal for the reactor core of a BWR. A control rod is X-shaped. Each vane of the X contains vertical metal tubes filled with boron powder. Boron is like neutron glue. When control rods are withdrawn from the reactor core, fewer neutrons released by fissioning atoms get absorbed by the boron inside control rods, leaving them free to cause other atoms to split. This increases the power level of the core. Conversely, when control rods are inserted, their boron absorbs neutrons to slow down, or even stop, the nuclear chain reaction.

Control rods are moved using water pressure acting upon pistons. Water pressure is applied to one side of the piston and vented from the other side to create the force necessary to move the control rod. By swapping which side of the piston receives water and gets vented, the resulting movement can cause the control rod to enter the reactor core or be withdrawn from it. During normal operation, differential pressure of approximately 250 to 300 pounds per square inch causes the control rod to move at a rate of about 3 inches per second, and it takes about 48 seconds for a control rod to travel the entire length from fully inserted to fully withdrawn or vice-versa. In an emergency, differential pressure of over 1,000 pounds per square inch can accelerate movement such that a control rod can move from fully withdrawn to fully inserted within 3 seconds.

The problem that GE Hitachi reported to the NRC in September 2010 involved increased control rod insertion times under emergency conditions because the control rods were slowed down by rubbing against the fuel channels around the fuel assemblies. As the fuel cell schematic on the preceding page illustrates, each fuel channel has two spacer buttons. The spacer buttons on adjacent fuel assemblies are designed to touch to keep space open for the control rod to move freely between them. But the control rods and fuel channels are made of metal that expands when heated. When not properly accommodated, this expansion can cause warping or bowing - the control rod can bend towards the fuel channels and/or the fuel channels can bend towards the control rod. When this occurs, the rubbing of the control rod against the fuel channels can slow it down or even stop it.

The initial problem GE Hitachi reported in September 2010 was limited to a certain type of control rod manufactured by GE Hitachi that was susceptible to rubbing under certain conditions. This initial problem was addressed by requiring owners of reactors equipped with these control rods to time the control rod movements more frequently to check for rubbing. The more frequent testing would continue until the problematic control rods could be replaced.

The September 2011 update to the initial report added a wrinkle that had not been previously considered.

Certain conditions, like an earthquake, could increase the likelihood that control rods would rub against channels. For example, shaking caused by an earthquake could cause fuel assemblies to twist and flex, narrowing or even eliminating the gap in which the control rods move. GE Hitachi reported the worst case scenario to be an event occurring when the pressure inside the reactor is below the normal operating pressure of 1,000 pounds per square inch.

October 3, 2011 Page 2 of 4

The water pressure used to move control rods in an emergency comes from two sources: (1) accumulators, and (2) the reactor vessel housing the reactor core itself. Each control rod is equipped with its own accumulator. An accumulator has two parts: a metal cylinder filled with water and a rounded tank pressurized with nitrogen gas. The picture on the left below shows two complete accumulators and the nitrogen tank from a third. The gauges at bottom indicate the pressures of the nitrogen inside the tanks.

The chart on the right above shows the time it takes a fully withdrawn control rod to insert into the reactor core during an emergency as a function of the reactor pressure. The yellow curve shows the insertion time when the differential pressure to move the control rod comes from only the accumulator. When the reactor pressure is low, the high pressure inside the accumulator rapidly inserts the control rod with little force to oppose it. As the reactor pressure increases (thus increasing the pressure holding the control rod out), it takes longer for the accumulator pressure to insert a control rod. The cyan (blue) curve shows the insertion time when the differential pressure to move the control rod comes from only the reactor vessel.

In this case, the insertion time shortens as the reactor pressure increases - exactly the opposite reaction from the accumulator case.

The accumulators are pressurized to 1,200 to 1,400 pounds per square inch (psi). Water from the accumulators cylinder is applied to one side of the control rod piston while the other side is vented to the atmosphere. The differential pressure across the piston drives the control rod into the reactor core. As the pressure inside the reactor vessel increases, the differential pressure across the piston remains the same (1,200 to 1,400 psi on one side and atmospheric on the other) but the control rod is moving against more force, which slows it down.

When the control rod is inserted using only pressure inside the reactor vessel, the differential pressure across the piston still causes that movement. But as the pressure inside the reactor vessel increases, so does the differential pressure across the piston (from around 500 psi to over 1,200 psi for the cyan curve in the chart). Although the resistance to control rod insertion also rises as the pressure inside the reactor vessel increases, the increase in differential pressure is the larger factor. This is due to the geometry of the piston itself - one side is significantly bigger than the other side. Consequently, equal force applied to October 3, 2011 Page 3 of 4

both sides of the piston causes it to insert because the force on one side acts upon a larger area and therefore has a greater effect.

In any case, emergency insertions of control rods occur when accumulator pressure and reactor pressure are both available. These two sources produce the red dashed curve in the chart. It shows that at low reactor vessel pressures, the accumulator is the dominant factor affecting control rod insertion times. As the pressure inside the reactor vessel rises, the accumulators influence diminishes until the reactor pressure becomes the dominant factor. The slowest control rod insertion time occurs at a reactor vessel pressure of around 800 psi, when neither source is dominant.

Returning to the safety problem GE Hitachi reported to the NRC on September 27, 2011, the problem is pronounced when the reactor vessel pressure is less than 900 psi.. This corresponds to the peak portion of the red dashed curve, and is when rubbing between control rods and fuel channels poses the greatest risk.

At higher reactor vessel pressures and at very low reactor vessel pressures, the control rod insertion is faster. The high differential pressures producing this speed are also most likely to overcome resistance caused by friction between the control rod and the fuel channels. But in that mid-range region where insertion times increase, friction can further slow or even stop control rod insertion.

Failure of control rods to fully insert can have disastrous consequences. The array of emergency systems that provide makeup water to the reactor vessel when a pipe breaks or other emergency occurs, as well as the massive concrete containment structures, are designed to mitigate an emergency once the control rods shut down the nuclear chain reaction. If the control rods fail to do so and the nuclear engine continues running, it may produce more energy than the emergency makeup and containment systems can handle.

When the problem first surfaced last fall, GE Hitachi recommended that plant owners increase testing of the control rod insertion times and apply a margin to the test results to account for rubbing between the control rod and fuel channels. With the recently expanded dimension of the problem (i.e., the fact that certain conditions can exacerbate the rubbing), GE Hitachi is recommending that plant owners increase the margin applied to the test results.

A better solution would be to design and use fuel bundles and control rods that did not rub against one another. Bump and grind is more suited for the dance floor than in the reactor core.

Prepared by: David Lochbaum Director, Nuclear Safety Project October 3, 2011 Page 4 of 4

Hello Chuck:

Attached is a November 2008 report to the NRC by the owner of the DC Cook nuclear plant in Michigan. Some turbine blades came apart. The turbine blades are very large pieces of metal rotating at least 1,800 times per minute. When they come apart, it ain't pretty.

As the blades came apart, they became missiles. Some missiles ripped through piping providing hydrogen gas to the generator, starting a fire. Other missiles ripped through piping containing lubricating oil for the turbine bearings, starting another fire.

No big deal right? The turbine/generator is not a safety-related component, so let it burn.

But check out the paragraph on page 3 of the Cook LER beginning with "At 2125 hours0.0246 days <br />0.59 hours <br />0.00351 weeks <br />8.085625e-4 months <br />, the fire water header...". The vibrations caused when heavy metal missiles struck concrete walls and floors created a mini-earthquake in the local area. As the ground shook, a buried fire protection header pipe ruptured. Water from the North Fire Water Storage Tank poured from the ruptured header. I've read other accounts where security officers in the vicinity reported "ol' faithful" as a geyser of water shot up from the ground.

Not good when an event triggering a fire also disables the fire suppression system. Not good at all.

The NRC's regulations require plants be designed to survive ground motion caused by the safe shutdown earthquake (SSE). As you've noted earlier, the SSE value for CGS is suspect. But for the moment, let's assume the SSE value is right.

The NRC's regulations do not require the entire plant to survive an SSE shake, only the parts and components of the plant necessary to shut down the reactor core and maintain it shut down.

Control rod system, yes. Fire protection system, no.

There's a disconnect between fire protection regulations and non-fire safety regulations. I spoke about that disconnect in a recent blog post (see http://allthingsnuclear.org/fire-safety-or-not/).

The Cook event was caused by a turbine blade failure. An SSE-magnitude quake could cause even greater ground motion at CGS. Would the SSE cause fire(s)? Would the SSE disable or degrade the fire protection systems?

Lubricating oil is not only used for the turbine. CGS has two very large recirculation pumps inside the drywell that used large amounts of lubricating oil. There's ample combustible material available in the reactor building.

It would be nice to know that CGS can survive an SSE even if that SSE starts a fire.

Thanks, Dave Lochbaum UCS