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Power Lines to Brunswick Nuclear Power Station in North Carolina
Oddly, Nuclear Power Stations depend on electricity from the grid for cooling of the reactor and its spent fuel pool. Above ground power lines at Brunswick nuclear power station, on the Atlantic Ocean, put it at high risk from hurricanes, as well as tornadoes. Building in background is Brunswick Nuclear Power Station. One can only hope that backup generators work, and that they can access enough diesel.

There are endless problems at aging nuclear power stations, which are reported by the US NRC using the euphemism of “event” like it’s a big nuclear party: http://www.nrc.gov/reading-rm/doc-collections/event-status/event/ Some “events” are indeed drunk or drugged up employees, euphemistically called “fitness for duty”, or rather lack thereof. It puts in mind those who hold parties during major hurricanes, rather than evacuating. While hurricane parties often end badly for the stupid participants, nuclear “events” could impact many innocents for generations.

A look at today’s auto industry, and all of the defects at nuclear power stations under construction, suggests that new nuclear power stations would not be safer either, and could be even less safe. This is apart from the fact that there is no known solution for the nuclear waste, and that they are permitted to discharge long-lived lethal radioactive materials into the environment as part of routine nuclear power station operations, as well as throughout the entire nuclear fuel chain. Would you rather a new, potentially defective, automobile, or an old one? Neither are safe for highway driving. And these are nuclear reactors.

And, researchers keep discovering that the risks of this ionizing radiation, which they are allowed to continual discharge into the environment, are worse than previously claimed by many. For those who missed it: https://miningawareness.wordpress.com/2015/12/19/another-look-at-the-recent-low-dose-radiation-exposure-study-inworks/ This recent 3 country, government funded, study suggests that excess cancer deaths, and thus cancer rates, are worse than even predicted by the US government funded, National Academy of Scienes, BEIR VII (2006) report. How much worse ranges from around three times worse than BEIR VII estimated to 26 times worse, (i.e. from per 100 people there would be 29 extra cancers per Sievert (1000 mSv) – which is approximately BEIR’s upper bound- to 262 or more extra cancers per Sv (1000 mSv) in a population of 100, meaning that 100% of the population having excess cancers would kick in at around 400 mSv cumulative exposure. The middle estimate interpretation is around 152 extra cancers per Sv (1000 mSv) meaning that a 100% excess cancer rate kicks in at around 700 mSv cumulative exposure. Cancer for everyone would occur at lower doses due to already high cancer rates. See: https://miningawareness.wordpress.com/2015/12/19/another-look-at-the-recent-low-dose-radiation-exposure-study-inworks/ Already in 1950, William Russell of Oak Ridge National (Nuclear) Lab stated regarding ionizing radiation: “There is no threshold dose. In other words, genetic changes may be expected at any dose, no matter how small…

Within the last week there have been nuclear reactor problems, which on their own could result in a nuclear meltdown. Combined they resemble the Three Mile Island nuclear disaster.

Dramatic headlines at Brunswick Nuclear Power Station in N. Carolina: “MANUAL SCRAM AND ALERT DECLARATION DUE TO ELECTRICAL FAULT RESULTING IN FIRE/EXPLOSION “. However, the more innocuous sounding “MANUAL REACTOR SCRAM FOLLOWING SPURIOUS OPENING OF TWO SAFETY RELIEF VALVES” at Perry Nuclear Power Station in Ohio may be more dangerous, but passes unnoticed, because less dramatic. Also operator “fitness” matters. If the operator is too tired or drugged up they may not act appropriately or in a timely manner. For most people, being drunk is a bad idea, as well, and can impact reaction. [Update: Subsequent to this “event” Perry suffered from a power failure, loss of cooling, too! Text added toward the bottom of the page.]

The Brunswick explosion and fire appears to have impacted the electricity needed to run the cooling water to keep the reactor (and spent fuel) from over-heating and melting down. An auxiliary transformer also failed. Luckily, or due to good maintenance, the backup generators operated. One of the most bizarre aspects of nuclear reactors is this dependence upon other sources of power – generally offsite electricity and thus their need to be connected to the grid. Alternatively, nuclear cooling systems can be powered with backup generators, which sometimes fail to start up and require diesel fuel to run. https://miningawareness.wordpress.com/2016/02/11/multiple-failures-dependence-upon-external-grid-led-to-near-miss-at-scottish-nuclear-power-station/ As explained by the US NRC: “The reactor’s core contains fuel assemblies that are cooled by water circulated using electrically powered pumps. These pumps and other operating systems in the plant receive their power from the electrical grid. If offsite power is lost, emergency cooling water is supplied by other pumps, which can be powered by onsite diesel generators. Other safety systems, such as the containment cooling system, also need electric power.http://www.nrc.gov/reactors/bwrs.html

In the Three Mile Island disaster, not only did an electrical or mechanical failure stop the pumps needed to cool the reactor, but a relief valve became stuck open, as it recently did at Perry Nuclear Power Station.

In the Three Mile Island Nuclear Disaster: “Either a mechanical or electrical failure prevented the main feedwater pumps from sending water to the steam generators that remove heat from the reactor core. This caused the plant’s turbine-generator and then the reactor itself to automatically shut down. Immediately, the pressure in the primary system (the nuclear portion of the plant) began to increase. In order to control that pressure, the pilot-operated relief valve (a valve located at the top of the pressurizer) opened. The valve should have closed when the pressure fell to proper levels, but it became stuck open. Instruments in the control room, however, indicated to the plant staff that the valve was closed. As a result, the plant staff was unaware that cooling water was pouring out of the stuck-open valve…” (USNRC summary)

Twenty years ago, the US NRC was already concerned about aging valves and valve operators: “NUREG/CR-6246 ORNL-6814
Effects of Aging and Service Wear on Main Steam Isolation Valves and Valve Operators , Manuscript Completed: February 1996 Date Published: March 1996
Prepared by R. L. Clark , Oak Ridge National Laboratory Managed by Lockheed Martin Energy Systems Oak Ridge National Laboratory Oak Ridge, TN 37831-6285 , J. Jackson, NRC Project Manager , Prepared for Division of Engineering Technology Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 NRC Job Code B0828

At Brunswick on Sunday, 7 Feb, 2016:
MANUAL SCRAM AND ALERT DECLARATION DUE TO ELECTRICAL FAULT RESULTING IN FIRE/EXPLOSION

At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode.

At 1704 EST the licensee reported the following:

“At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves.

“The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control.

“The Plant response to the event was per design.
“Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT.
“The public health and safety is not impacted by this event.”
At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a “loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes.”
The NRC Resident Inspector has been notified.
The licensee has notified the State and Local governments.
Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).
* * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * *
At 1814 EST the emergency declaration was terminated because offsite power was restored. / The NRC Resident Inspector has been notified.
The licensee has notified the State and Local governments.
Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160208en.html

At Brunswick Nuclear Power Station there have been valve problems in the past:
Manual Reactor Scram Due to Spurious Safety Relief Valve Opening…. On November 9, 2008, at 1108 hours Eastern Standard Time (EST), Safety Relief Valve (SRV) 2-B21-F013H [SB] spuriously opened with no Operator action or testing in progress…
‘EVENT CAUSE
The root cause of this event was the failure to verify proper seating of the set pressure spring in the upper spring follower plate. The set pressure spring was installed with part of the spring sitting on the ledge of the upper spring follower plate, which actually compressed the spring more than it would when properly installed. The SRV H pilot stage assembly was then certified with the spring incorrectly positioned on the ledge. The BSEP corrective maintenance procedure, OCM-VSR509, “Main Steam Relief Valves Target Rock Model 7567 Air Operators and Pilot Assembly Disassembly, Inspection, and Reassembly,” did not provide guidance to check or verify that the spring is properly installed. It should be noted that prior to 2000, Target Rock performed rebuilds of BSEP’s SRV pilot stage assemblies. In mid-2000, BSEP Maintenance began performing the rebuilds of the SRV pilot stage assemblies.

https://lersearch.inl.gov/PDFView.ashx?DOC::3242008002R00.PDF

On Monday, February 8, 2016:
Facility: PERRY, Region: 3 State: OH, Unit: [1] [ ] [ ]. RX Type: [1] GE-6
NRC Notified By: MICHAEL DOTY. HQ OPS Officer: DONALD NORWOOD
Notification Date: 02/08/2016. Notification Time: 17:50 [ET]. Event Date: 02/08/2016
Event Time: 15:03 [EST]. Last Update Date: 02/08/2016
Emergency Class: NON EMERGENCY
10 CFR Section: 50.72(b)(2)(iv)(B) – RPS ACTUATION – CRITICAL
Person (Organization): DAVID HILLS (R3DO)
Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 100 Power Operation 0 Hot Shutdown
Event Text
MANUAL REACTOR SCRAM FOLLOWING SPURIOUS OPENING OF TWO SAFETY RELIEF VALVES
“At 1500 EST on February 8, 2016, two safety relief valves (SRV) opened upon a spurious Division 2 initiation signal. This caused suppression pool temperature to increase. At 1503 EST, plant operators took action to manually SCRAM the reactor at 95 degrees Fahrenheit in the suppression pool per plant procedures. The SRVs closed immediately following the scram at 1503 EST. The cause of the SRVs opening is currently under investigation.
“During the scram, all rods fully inserted into the core. Reactor Pressure is stable with decay heat being removed via turbine bypass valves to the main condenser. Reactor level control is currently being maintained via feedwater. Main Steam Isolation Valves are open. Cool down and depressurization to Mode 4 to follow. The plant is in a normal post SCRAM electrical line-up.”
The licensee notified the NRC Resident Inspector.
http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160209en.html

[Update: Subsequently, another event!
Event Number: 51729, Facility: PERRY, Region: 3 State: OH, Unit: [1] [ ] [ ], RX Type: [1] GE-6
NRC Notified By: TONY KLEDZIK, HQ OPS Officer: DONALD NORWOOD
Notification Date: 02/11/2016, Notification Time: 19:38 [ET]
Event Date: 02/11/2016, Event Time: 15:04 [EST]
Last Update Date: 02/11/2016, Emergency Class: NON EMERGENCY
10 CFR Section: 50.72(b)(3)(iv)(A) – VALID SPECIF SYS ACTUATION
Person (Organization): DAVID HILLS (R3DO)
Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N N 0 Cold Shutdown 0 Cold Shutdown
Event Text
AUTOMATIC START OF EMERGENCY DIESEL GENERATOR AND LOSS OF SHUTDOWN COOLING
“At 1504 EST on February 11, 2016, with the plant shutdown in a forced outage, the Division 1, 4.16 Kv Safety Bus (EH11) lost power. Division 1 Shutdown Cooling was in service at the time and the Division 1 Shutdown Cooling pump A tripped. The Division 1 Emergency Diesel Generator (EDG) started and loaded EH11 as designed. However, the Emergency Service Water (ESW) A pump, which supplies cooling water to the EDG did not start. Due to the absence of cooling water to the EDG, operators took manual action to secure the Division 1 EDG. Division 2 Shutdown Cooling was operable during this transient and was subsequently started. The Division 1 Shutdown Cooling common suction isolation valve (1E12F0008) had previously been de-energized in the open position to support planned maintenance. The Division 2 Shutdown Cooling isolation valve was not affected by the loss of bus EH11. Shutdown Cooling was re-established at 1544 EST using the Division 2 Shutdown Cooling pump. Reactor coolant temperature rose from approximately 89 degrees Fahrenheit to 115 degrees Fahrenheit during the event. The cause of the loss of EH11 and subsequent failure of ESW A pump to start are currently under investigation.

“This event is being reported under 10 CFR 50.72(b)(3)(iv)(A) as a specific system actuation due to the auto start of the Division 1 EDG on a valid signal.

“The plant remains shutdown with Division 2 Shutdown Cooling in operation. The plant is in a normal electrical line up with the exception of bus EH11 being de-energized.”

The licensee notified the NRC Resident Inspector.
http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2016/20160212en.html ]

Drunk or drugged up employees remain another problem at US Nuclear Power Stations. At River Bend Nuclear Power Station in Louisiana: “A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee’s access to the plant has been terminated”. At Brunswick: “POSITIVE RANDOM FITNESS FOR DUTY TEST RESULT A contract employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee’s access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2015/20150316en.html At McGuire Nuclear Power Station: “This letter refers to an Event Notification (EN) 50631 made by McGuire Nuclear Station on November 20, 2014, to the U.S. Nuclear Regulatory Commission (NRC). This EN informed the NRC that you tested positive for an illegal drug during a random fitness-for-duty (FFD) test conducted on November 17, 2014. … we were informed that McGuire Nuclear Station no longer had a need to maintain your senior operator license’http://pbadupws.nrc.gov/docs/ML1506/ML15065A233.pdf

The US NRC short version of the Three Mile Island Nuclear Disaster:
The accident began about 4 a.m. on Wednesday, March 28, 1979, when the plant experienced a failure in the secondary, non-nuclear section of the plant (one of two reactors on the site). Either a mechanical or electrical failure prevented the main feedwater pumps from sending water to the steam generators that remove heat from the reactor core. This caused the plant’s turbine-generator and then the reactor itself to automatically shut down. Immediately, the pressure in the primary system (the nuclear portion of the plant) began to increase. In order to control that pressure, the pilot-operated relief valve (a valve located at the top of the pressurizer) opened. The valve should have closed when the pressure fell to proper levels, but it became stuck open. Instruments in the control room, however, indicated to the plant staff that the valve was closed. As a result, the plant staff was unaware that cooling water was pouring out of the stuck-open valve.

As coolant flowed from the primary system through the valve, other instruments available to reactor operators provided inadequate information. There was no instrument that showed how much water covered the core. As a result, plant staff assumed that as long as the pressurizer water level was high, the core was properly covered with water. As alarms rang and warning lights flashed, the operators did not realize that the plant was experiencing a loss-of-coolant accident. They took a series of actions that made conditions worse. The water escaping through the stuck valve reduced primary system pressure so much that the reactor coolant pumps had to be turned off to prevent dangerous vibrations. To prevent the pressurizer from filling up completely, the staff reduced how much emergency cooling water was being pumped in to the primary system. These actions starved the reactor core of coolant, causing it to overheat.

Without the proper water flow, the nuclear fuel overheated to the point at which the zirconium cladding (the long metal tubes that hold the nuclear fuel pellets) ruptured and the fuel pellets began to melt. It was later found that about half of the core melted during the early stages of the accident. Although TMI-2 suffered a severe core meltdown, the most dangerous kind of nuclear power accident, consequences outside the plant were minimal. Unlike the Chernobyl and Fukushima accidents, TMI-2’s containment building remained intact and held almost all of the accident’s radioactive material.

Federal and state authorities were initially concerned about the small releases of radioactive gases that were measured off-site by the late morning of March 28 and even more concerned about the potential threat that the reactor posed to the surrounding population. They did not know that the core had melted, but they immediately took steps to try to gain control of the reactor and ensure adequate cooling to the core. The NRC’s regional office in King of Prussia, Pa., was notified at 7:45 a.m. on March 28. By 8 a.m., NRC Headquarters in Washington, D.C., was alerted and the NRC Operations Center in Bethesda, Md., was activated. The regional office promptly dispatched the first team of inspectors to the site and other agencies, such as the Department of Energy and the Environmental Protection Agency, also mobilized their response teams. Helicopters hired by TMI’s owner, General Public Utilities Nuclear, and the Department of Energy were sampling radioactivity in the atmosphere above the plant by midday. A team from the Brookhaven National Laboratory was also sent to assist in radiation monitoring. At 9:15 a.m., the White House was notified and at 11 a.m., all non-essential personnel were ordered off the plant’s premises.

By the evening of March 28, the core appeared to be adequately cooled and the reactor appeared to be stable. But new concerns arose by the morning of Friday, March 30. A significant release of radiation from the plant’s auxiliary building, performed to relieve pressure on the primary system and avoid curtailing the flow of coolant to the core, caused a great deal of confusion and consternation. In an atmosphere of growing uncertainty about the condition of the plant, the governor of Pennsylvania, Richard L. Thornburgh, consulted with the NRC about evacuating the population near the plant. Eventually, he and NRC Chairman Joseph Hendrie agreed that it would be prudent for those members of society most vulnerable to radiation to evacuate the area. Thornburgh announced that he was advising pregnant women and pre-school-age children within a five-mile radius of the plant to leave the area.
http://www.nrc.gov/reading-rm/doc-collections/fact-sheets/3mile-isle.html

Other investigations have concluded that, while equipment failures initiated the event, the fundamental cause of the accident was “operator error.”… While we agree that this statement is true, we also feel that it does not speak to the fundamental causes of the accident. Let us consider some of the factors that significantly contributed to operator confusion…” President’s Commission on Three Mile Island: http://www.threemileisland.org/downloads/188.pdf

Easy to understand information by expert David Lochbaum: http://allthingsnuclear.org/dlochbaum/when-safety-relief-valves-fail-to-provide-safety-or-relief-at-nuclear-plants http://www.ucsusa.org/sites/default/files/legacy/assets/documents/nuclear_power/20071001-harris-ucs-brief-transformer-event.pdf

Longer version of Brunswick valve incident in 2008: “Brunswick Steam Electric Plant (BSEP), Unit 2 05000324, Manual Reactor Scram Due to Spurious Safety Relief Valve Opening…. On November 9, 2008, at 1108 hours Eastern Standard Time (EST), Safety Relief Valve (SRV) 2-B21-F013H [SB] spuriously opened with no Operator action or testing in progress. The SRV’s control switch was cycled as required by Abnormal Operating Procedure with no success. At 1113 hours the fuses were pulled for SRV H in an attempt to close the valve. At 1117 hours, a manual reactor scram was inserted based on the Suppression Pool temperature reaching 109.8 degrees Fahrenheit (F). Technical Specifications requires a manual reactor scram to be inserted when Suppression Pool average temperature exceeds 110 degrees F. All control rods fully inserted from the manual reactor scram signal. Reactor water level lowered to Low Level 2 resulting in Primary Containment Isolation System (PCIS) [JM] isolations of Groups 2, 3, 6, and 8. In addition, the Reactor Core Isolation Cooling (RCIC) [BN] system actuated and injected into the reactor. The High Pressure Coolant Injection (HPCI) [BJ] system actuated but did not inject since reactor water level had recovered. An Alternate Rod Insertion [JC] signal was received, the Standby Gas Treatment (SBGT) [BH] system initiated, and the Reactor Recirculation [AD] pumps tripped as designed. All systems responded as designed.

At 1208 hours (EST) on November 9, 2008, the NRC was notified of this event (i.e., Event Number 44647) in accordance with 10 CFR 50.72(b)(2)(iv)(B), as an event or condition that results in actuation of the reactor protection system (RPS) [JC] when the reactor is critical, and 10 CFR 50.72(b)(3)(iv)(A), as an event or condition that results in valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B).

This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A), as an event or condition that resulted in manual or automatic actuation of any of the systems listed in 10 CFR 50.73(a)(2)(iv)(B)… EVENT CAUSE

The root cause of this event was the failure to verify proper seating of the set pressure spring in the upper spring follower plate. The set pressure spring was installed with part of the spring sitting on the ledge of the upper spring follower plate, which actually compressed the spring more than it would when properly installed. The SRV H pilot stage assembly was then certified with the spring incorrectly positioned on the ledge. The BSEP corrective maintenance procedure, OCM-VSR509, “Main Steam Relief Valves Target Rock Model 7567 Air Operators and Pilot Assembly Disassembly, Inspection, and Reassembly,” did not provide guidance to check or verify that the spring is properly installed. It should be noted that prior to 2000, Target Rock performed rebuilds of BSEP’s SRV pilot stage assemblies. In mid-2000, BSEP Maintenance began performing the rebuilds of the SRV pilot stage assemblies.https://lersearch.inl.gov/PDFView.ashx?DOC::3242008002R00.PDF

Fatigue Exemption Hurricanes: http://pbadupws.nrc.gov/docs/ML1111/ML11110A021.pdf
McGuire Fitness for Duty (not): http://pbadupws.nrc.gov/docs/ML1506/ML15065A233.pdf
McGuire Fitness for Duty (not)”event” due to drugs:
http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2014/20141121en.html

Many thanks to Mary of the Flying Cuttlefish Blog-Louisiana Sinkhole for an emailed alert. Here is the FC Report: https://flyingcuttlefish.wordpress.com/2016/02/09/brunswick/