Drought: How Much Drier Will It Get In Our Area?

This is not good for the beginning of summer.

USGS Current Water Data for the Nation

We’ve had a lot of spectacularly beautiful days recently. Today is one of those days?

Drought Monitor

Mid-Atlantic and Northeast

Dry conditions prevailed over much of the region, though well-placed showers (1-2 inches) in southwestern Pennsylvanian and environs led to the removal of Moderate Drought (D1) and a reduction of Abnormal Dryness (D0) across the central Appalachians. In contrast, D0 and D1 were increased from New York into New England due to declining streamflows (locally below the 10th percentile) and a lack of rain over the past 90 days (less than half of normal). In fact, many of the Northeast’s D1 areas are now running rainfall deficits in excess of 6 inches over the past 6 months.

Part 21 Notification of Junk Breaker and Electric Control Devices

Basically, this comes out of the River Bend and other plant concerns with large breaker reliability. Look at the  notification with the *.

I think this is unprecedented?

* 2016-31-00	Curtiss-Wright Nuclear Division	Interim Report Regarding a Nonconformance on Struthers-Dunn Relay Part No. B255XCXPFHSC125V Supplied to PSEG	06/08/2016	ML16162A679
* 2016-29-00	ABB Inc.	Notification of Deviation Related to K-Line Circuit Breaker Secondary Trip Latch	06/03/2016	ML16162A681
* 2016-28-00	Xcel Energy Prairie Island	Initial Notification of a Failure to Comply Related to ABB Circuit Breaker Type K-600S EO	06/03/2016	ML16162A682
			06/02/2016	ML16154A268
*2016-26-01	Ametek Solidstate Controls	Notification of Potential Defect with Ametek Inverter Manufactured with Signal Transformer R-10607 (Updated)	06/02/2016	ML16162A683
* 2016-26-00	Ametek Solidstate Controls	Notification of Potential Defect with Ametek Inverter Manufactured with Signal Transformer R-10607	06/02/2016	ML16162A680
* 2015-26-02	AZZ/Nuclear Logistics Inc.	Potentially Unqualified Component in Certain Allen Bradley Model 700RTC Timing Relays	05/26/2016	ML16154A207
* 2016-24-00	AZZ/Nuclear Logistics Inc.	Potential Reportable Condition Related to Eaton Freedom Series Contactor	05/19/2016	ML16154A205
2016-23-00	Nutherm International, Inc.	Potential Defect Found in a Moore Industries SCT Series Signal Converter	05/17/2016	ML16154A206
* 2016-22-00	United Controls International	Thomas & Betts Power Solutions/Cyberex Printed Circuit Boards and Mersen (formerly Ferraz Shawmut) Fuses	05/16/2016	EN 51936
* 2016-20-01	AZZ/Nuclear Logistics Inc.	Masterpact NT and NW Style Circuit Breakers Failed to Electrically Close Following an Anti-Pump Condition (Updated)	05/13/2016	ML16139A832
* 2016-20-00	AZZ/Nuclear Logistics Inc.	Masterpact NT and NW Style Circuit Breakers Failed to Electrically Close Following an Anti-Pump Condition	05/12/2016	ML16139A833
* 2016-19-00	Electroswitch	Part 21 Notification on Various Electroswitch Products Sold as Safety Class 1E	05/10/2016	ML16139A834
* 2016-11-01	Rotork Controls, Inc.	Part 21 Notification Concerning V12 [Part No N69-921] and K5 [N69-838 & N69·926] Safety Related Micro Switches	05/04/2016	ML16127A595
* 2016-16-01	National Technical Systems	Updated Potential Part 21 on Siemens 401-158 Safety Clip used on Type 3AF Circuit Breakers	04/29/2016	ML16126A111
2015-88-01			04/26/2016	ML16118A372
* 2016-14-00	Prairie Island Units 1 and 2	Interim Report of a Deviation or Failure to Comply Related to a Load Sequencer Undervoltage Relay	04/14/2016	ML16106A207
* 2015-26-01	AZZ / Nuclear Logistics	Potentially Unqualified Component in Certain Allen Bradley Model 700RTC Timing Relays (Update)	04/08/2016	ML16109A177
2016-07-01	Louisiana Energy Services	Final Report for Potential 10 CFR 21 Notification	04/06/2016	ML16104A403
			03/22/2016	ML16102A078
* 2016-13-00	United Controls International	Part 21 Notification for Mersen (formerly Ferraz Shawmut) OT15 Fuses	03/22/2016	ML16098A446
2015-20-01	United Controls International	Follow Up to Resistive Short Identified on Thomas & Betts Power Solutions/Cyberex Sense & Transfer Module	03/22/2016	ML16098A445
*2016-11-00	Rotork Controls, Inc.	Interim Report Related to a Basic Micro Switch Which Did Not Change State	03/18/2016	ML16088A087
*2016-16-00	National Technical Systems	Potential Part 21 on Siemens 401-158 Safety Clip used on Type 3AF Circuit Breakers

Indian Point Junk

They just finished big old outage and they got a crack in the 20 inch service water system. Any dummy can tell they don't have a process to keep their service water system pipes clean of dangerous cracks and flaws. it become a patchwork process of dealing with one crack at a time and not a system of making the piping system durable enough to not interrupt plant operations.

I wonder if they didn't have the baffle bolt problem would they even shutdown. 

Remember the snap back from Brexit?

Facility: INDIAN POINT Region: 1 State: NY Unit: [2] [ ] [ ] RX Type: [2] W-4-LP,[3] W-4-LP NRC Notified By: JUSTIN MACDONALD HQ OPS Officer: BETHANY CECERE	Notification Date: 06/24/2016 Notification Time: 04:05 [ET] Event Date: 06/24/2016 Event Time: 04:00 [EDT] Last Update Date: 06/24/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(i) - PLANT S/D REQD BY TS	Person (Organization): GLENN DENTEL (R1DO) CHRIS MILLER (NRR) WILLIAM GOTT (IRD)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
2	N	Y	93	Power Operation	92	Power Operation

Event Text

TS REQUIRED SHUTDOWN DUE TO LEAKING SERVICE WATER WELD ON CCW HX "At 0400 (EDT) on June 24, 2016, Indian Point Unit 2 initiated actions to commence reactor shutdown to comply with Technical Specification (TS) LCO 3.7.7, Condition B. TS LCO 3.7.7, Condition A had been entered at 0230 on June 21, 2016 in order to repair a leaking weld on the 20 inch service water pipe to nozzle weld on the 21 Component Cooling Water Heat Exchanger (CCW HX). Condition A allows 72 hours to restore the inoperable CCW train to service or Condition B is entered which requires the plant to be in Mode 3 in 6 hours and Mode 4 in 12 hours. "The initiation of a nuclear plant shutdown required by TS requires a 4-hour report in accordance with 50.72(b)(2)(i) which is being made by this notification." The licensee notified the New York Independent System Operator and the New York Public Service Commission. The licensee notified the NRC Resident Inspector.

Junk Plant Hatch Plant Junk SRVs

05000321/366

This is the 3 stage model that got Pilgrim plant into so much trouble last year. This model is supposed to be immune to setpoint drift. This is why they recently jumped out of the 2 stage SRVs.

I think Target Rock thinks why even bother playing with these little nuke boys. Or the valves are to old to repaired and tested by them. So the local two bit "NWS Technologies" now does the testing and repair on the obsolete SRVS.  

Across the board, Hatch has been having a lot of problems with maintaining the reliability of the 2 stage or the three stage SRVs for many years. They wasted a lot of money on their SRVS.

I believe the dimensions and the material quality of the components are in play. We generially don't know the true material quality of the component, there is just is no facts to predict how these components will fail.       

Edwin I. Hatch Nuclear Plant Unit 1

LER 2016-004-001
Safety Relief Valves As Found Settings Resulted in Not Meeting Tech Spec
Surveillance Criteria

On March 30 2016, with Unit 1 at 100 percent rated thermal power (RTP) , "as-found" testing of the 3-stage main steam safety relief valves (SRVs) (EllS Code RV) showed that two of the eleven main steam SRVs that were tested had experienced a drift in pressure lift setpoint during the previous operating cycle such that the allowable technical specification (TS) surveillance requirement (SR) 3.4.3.1 limit of 1150 +1- 34.5 ( +1- 3%) psig had been exceeded. Below is a table illustrating the Unit 1 SRVs that failed as found testing results after being removed from service during the Spring

2016 refueling outage.
MPL
1821-F013D
1B21-F013E 

Usually it's seat/valve bonding. This gap thing problem is new. In the Pilgrim 3 stage issue, they seemed to blame it on inappropriate test stand testing. I always thought the components in the valves are of a poor quality. There is generially poor service from Target Rock. I don't trust these guys and I don't trust the NRC's diagnoses with why the failing and why target rock is getting to unreliable.  

The SRV pilots were disassembled and inspected while investigating the reason for the drift. SNC has determined that the abutment gap closed pre-maturely. The pre-mature abutment gap closure is most likely due to loose manufacturing tolerances leading to SRV setpoint drift. They assume they know what caused this.

I am pretty sure if they were up at power with two SRVs inop, then they would be immediately be required to shutdown the plant.  

The two SRVs which failed to meet their Tech Spec required actuation pressure setpoint lifted early (3.2% low and 3.8% low). None of the eleven SRVs tested this cycle had as-found test results out of range high. Therefore, since the two identified SRVs lifted earlier than expected, the ASME Code Limit of 1375 psig peak vessel pressure would be maintained under normal and accident conditions. The opening of one or more SRVs at lower pressures would result in a less severe transient with reduced peak vessel pressure. Also, the slightly lower actuating pressure does not pose a significant LOCA initiator threat because the reactor steam dome would not experience > 11 00 psig during normal operation.

It is utterly disgraceful they can't detect these problems before installation in the plant.

The vendor specifications will be revised to tighten as-left tolerances of abutment and pre-load gap, increase the minimum set for abutment pressure at the high end of specification, and tighten diametrical and face run-out tolerances for bellows assembly on pre-load spacer mounting end.

Junk New Plant Watts Bar Has Another Scram

The first scram accrued on June 5. Did TVA throttle their employees from discovering these problems before start-up?  

Power Reactor	Event Number: 52026
Facility: WATTS BAR Region: 2 State: TN Unit: [ ] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: MATTHEW MILLER HQ OPS Officer: DANIEL MILLS	Notification Date: 06/20/2016 Notification Time: 17:19 [ET] Event Date: 06/20/2016 Event Time: 15:40 [EDT] Last Update Date: 06/20/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL 50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION	Person (Organization): BINOY DESAI (R2DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
2	A/R	Y	32	Power Operation	0	Hot Standby

Event Text

AUTOMATIC REACTOR TRIP "On June 20, 2016 at 1540 EDT, Watts Bar Nuclear Plant Unit 2 reactor tripped due to [reaching the] automatic Lo-Lo steam generator trip [setpoint] on [the] #4 steam generator. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. "All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. The station is in a normal shutdown electrical alignment. "The cause is currently under investigation. "This is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72 (b)(2)(iv)(B). "The NRC Senior Resident has been notified." There was no effect on Unit 1.

Junk Plant Pilgrim: Why Did They Shutdown Yesterday?

Update 6/23

Now at 30%

Correction, I misread the list. Salem and Pilgrim are right near each other. I was reading the Salem power level as Pilgrim. Sorry.

Pilgrim was always at 100% power.


????

Indian Point 2 Is Beginning To Make Power

Where is the outrage by the governor: I think they are all in cahoots with each other.

Why is River Bend and Pilgrim shutdown?

Indian Point 2

2%

Coordinated Attack On Israel and USA?

Basically 3 days later? 

June 8 : "The military wing of Palestinian terror group Hamas called the Muslim holy month of Ramadan the “month of jihad,” in an article published on Tuesday — a day before two West Bank terrorists killed four people in an attack in Tel Aviv"

Should we bulldoze his parents house like they do in Israel??? 

Junk UCS: Maybe Dave Should Retire?

I think Dave is captured by the NRC. He is trading access for a independent description and the anticipation of future conditions in the nuclear industry.

UCSat NRC Region II: Downsizing without Compromising Nuclear Safety Margins

Maybe Dave is giving himself green across the board for the access the NRC gives him?

We in a historic financial crisis in the industry. They are cutting back funding across the board willy nilly.  The republicans are attacking the NRC as never before. They are blackmailing by threatening budget cuts, the NRC into reducing oversight. What do you think about de-coupling the LOCA from the LOOP?   

I give the NRC a red finding with their inability to control the nuclear industry. 

The UCS gets a yellow finding...

NRC Is A Power Unto Themselves (Uncontestable): What The Court Said

I understand the NRDC doesn't speak for the interest of us all and they wrote up the contention for their own agenda.

It is beyond chilling in a big way, where the court are too afraid to criticize congresses handling of nuclear waste on the big picture.

Big picture, I think they are saying congress set up regulation over nuclear power power beyond the preview of the courts. If congress runs the nuclear power federal oversight as a candy jar, the courts have no power to maintain order with the NRC and the industry. The courts have no power to make the nuclear power to serve the greater nation ends or public ends. Fundamentally the republicans and businesses have weakened the independence and separation of  three branches of government...they have drastically weakened the power of the courts to serve the public good. This is what the controversy with the philosophy of a activist courts. Gaming the activist court issue...the courts can't make at independent interpretation of the congressional rules on the greater interest of the peoples interest, but you can go full steam ahead with court activism if its in the interest of the elites and business interest. Its as if  the republicans, the elites, the businesses, through congress, has captured Congress and the presidency.  

I get it our Constitution sets up our government with three seperate and independent branches...but the courts still have the power to keep (supposed) the politicians clean. I see this this court as too timid to make political waves.

And most chilling of all, our opinion of congress is at historic lows. It is as our votes don't count and the courts don't care.  

Vermont appeal of NRC rule shot down by court By Robert Audetteraudette@reformer.com @audette.reformer on Twitter Posted:   06/09/2016 11:08:21 AM EDT | Updated:   about 10 hours ago Click photo to enlarge A federal appeals court in Washington, D.C., shot down a petition filed by... (Reformer file photo) « 1 » BRATTLEBORO >> If the states and the National Resources Defense Council are unhappy with regulations promulgated by the Nuclear Regulatory Commission, then take it up with Congress. That was the conclusion of the Court of Appeals for the District of Columbia to an appeal lodged by several attorneys general, environmental organizations and one Native American community about the NRC's spent fuel handling and storage regulations. "We acknowledge the political discord surrounding our nation's evolving nuclear energy policy," wrote the court. "But the role of Article III courts in this debate is circumscribed." The scope of review under the arbitrary and capricious standard is narrow and a court is not to substitute its judgment for that of the agency, stated the decision, rendered on June 3. "To the extent that the petitioners disagree with the NRC's current policy for the continued storage of spent nuclear fuel, their concerns should be directed to Congress." The appeal contended that the NRC utilized "several unreasonable assumptions," including that spent nuclear fuel will be removed from spent-fuel pools within 60 years of reactor decommissioning; that after the 60-year period, spent fuel will be stored in dry casks that are replaced every 100 years; and that institutional controls over spent nuclear fuel will exist into perpetuity. "We hold that none of these assumptions is so unreasonable as to render the NRC's decision-making arbitrary or capricious," noted the court. "We therefore deny the petitions for review on this issue." The NRC deserves "deference" in its decisions, wrote the court, because "An agency does not engage in arbitrary or capricious decision-making by making 'predictive judgment(s)' or even by relying on '(i)ncomplete data.'"

Nuts

PUBLIC MEETING ANNOUNCEMENT

Title: Meeting to Discuss the Need for a Rule for Risk Informed Decoupling of Assumed Loss-of-
Offsite Power from Loss-of-Coolant Accident Analysis

Date(s) and Time(s): June 28, 2016, 01:00 PM to 02:00 PM

Updated:There is a hell of reduction in the burdens with supporting safety. You couple the LOCA and LOOP together because of the enormus complexity of the system. This makes you have a sense that the system might not respond as intended. I seen it personally in Vermont Yank's LOOP in 1992, you can see many times in all  the Pilgrim LOOPs  and in the recent dual plant Millstone. 

Who is going to speak to all of the problems in troublesome LOOPs in this presentation? 

Meeting to Discuss the Need for a Rule for Risk Informed Decoupling of Assumed Loss-ofOffsite Power from Loss-of-Coolant Accident Analysis Date(s) and Time(s): June 28, 2016, 01:00 PM to 02:00 PM Location: NRC One White Flint North, O 9B4 11555 Rockville Pike Rockville, MD Category: This is a Category 3 meeting. Public participation is actively sought for this meeting to fully engage the public in a discussion of regulatory issues. Purpose: Provide an opportunity for external stakeholders and the NRC staff to exchange information on the need for a rulemaking action for Risk Informed Decoupling of Assumed Loss-of-Offsite Power from Loss-of-Coolant Accident Analysis and related petition for rulemaking (PRM) 50-77

Contact: Robert H. Beall 301-415-3874 Robert.Beall@nrc.gov NRC NRC Staff Participants: External NEI representatives Teleconference: Interested members of the public can participate in this meeting via a toll-free teleconference. For details, please call the NRC meeting contact.

Junk Plant Susquehanna Leak Violations: At Least A Red finding?

Is this Susquehanna’s version of Indian Point’s baffle bolt cracks problem???

Basically the NRC is implicated in this outcome. They weren't assertive enough to call for a immediate showdown. We don't know the role of the agency play with their complicity in not making Susquehanna detect the flaw before the leak showed up. It is now a classic conflict of interest with them setting up the violation because they were complicit with their non participation. They need to get a outsider (outside the NRC)to inspect this violation and then set the violation level. All they are doing is basically grading themselves.



May 9, 2016 @ 11 am : Updated and now with a picture of the LPRM tubes.

A month from the big maintenance refueling outage and they got two reactor water leaks???  

• BERWICK, Pa., March 12, 2016 /PRNewswire/ -- Operators at Talen Energy's Susquehanna nuclear power plant disconnected the Unit 1 reactor from the electrical grid late Friday, March 11 into early Saturday, March 12  to begin a scheduled refueling and maintenance outage.

It is really hot under the reactor. More so if they have a lot of fuel failure. This will be the way most US plant expire...basically very high and preventable very high radiation levels will make it too expensive to keep the plants on the grid.

The area under the core is extraordinarily radioactive and contaminated. We had to dress up in triple anti C clothing, a rain suit and respirators. With all the tubes, CDRMs and wires directly under the bottom reactor head, it looked like a up-side-down rain forest, if you had a imagination. It was constantly leaking astonishing highly radioactive water down on us. If they gave us a job down there, they'd say mockingly, "buddy, you are going down to our rain forest" with a huge smile.      

Susquehanna is a very busy two plant. They got problems with too many plant scrams. Here is another blatant example of dangerous certainty/uncertainty gaming to override safety rules. It’s basically Davis Besse redux. It’s as if nobody in the industry has learned a thing about pressure barrier leaks and Davis Besse.

REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH REGULATORY GUIDE 1.45 (Draft was issued as DG-1173, dated June 2007) 

GUIDANCE ON MONITORING AND RESPONDING TO REACTOR COOLANT SYSTEM LEAKAGE

A. INTRODUCTION 

This revision to Regulatory Guide 1.45 (Ref. 1) describes methods that the staff of the U.S. Nuclear Regulatory Commission (NRC) considers acceptable for use in implementing the regulatory requirements specified below with regard to selecting reactor coolant leakage detection systems, monitoring for leakage, and responding to leakage. This guide applies to light-water-cooled reactors.

General Design Criterion (GDC) 14, “Reactor Coolant Pressure Boundary,” as set forth in Appendix A, “General Design Criteria for Nuclear Power Plants,” to Title 10, Part 50, “Domestic Licensing of Production and Utilization Facilities,” of the Code of Federal Regulations (10 CFR Part 50), (Ref. 2), requires that licensees or applicants design, fabricate, erect, and test the reactor coolant pressure boundary (RCPB) so as to ensure an extremely low probability of abnormal leakage, rapidly propagating failure, and gross rupture. As a result, the design of these nuclear components normally follows the criteria established in Section III of the Boiler and Pressure Vessel Code (Ref. 3) promulgated by the American Society of Mechanical Engineers (ASME). 

During the design phase, degradation-resistant materials are normally specified for reactor coolant system (RCS) components. However, materials can degrade as a result of the complex interaction of the materials, the stresses they encounter, and the normal and upset operating environments they experience. Such material degradation could lead to leakage of the reactor coolant. Consequently, GDC 30, “Quality of Reactor Coolant Pressure Boundary” (Ref. 2), requires that plants provide the means for detecting and, to the extent practical, identifying the location of the source of reactor coolant leakage. 

(ASME: Basically a private regulator writing up corporate sponsored engineering codes from money.)

Basically the containment has rudimentary instrumentation with detecting leaks in the containment. You get it, basically the first leak obscured the control room indications of the second leak? There is no way to detect the second leak. Do you know what a ghost containment leak is? Its having a history of a bunch of non pressure barrier or a prolonged non pressure barrier leaks in containment. This imprints in the minds of the control operators. Months or years later, a pressure barrier leak emerges. The control room operators then assume the new leak is coming from the same component that has past leakage problems. This is Davis Besse and TMI.

Power Reactor	Event Number: 51987
Facility: SUSQUEHANNA Region: 1 State: PA Unit: [1] [ ] [ ] RX Type: [1] GE-4,[2] GE-4 NRC Notified By: LONNIE CRAWFORD HQ OPS Officer: HOWIE CROUCH	Notification Date: 06/08/2016 Notification Time: 07:01 [ET] Event Date: 06/08/2016 Event Time: 02:26 [EDT] Last Update Date: 06/08/2016
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(A) - DEGRADED CONDITION	Person (Organization): FRANK ARNER (R1DO)

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

DISCOVERY OF UNISOLABLE REACTOR PRESSURE BOUNDARY LEAKAGE "Susquehanna Unit 1 identified RPV [reactor pressure vessel] pressure boundary leakage from [local power range monitor] LPRM 24-09 housing above the flange during an under vessel leak inspection on 06/08/2016 at 0226 EDT. The leakage point is a through wall indication on the ASME Class 1 LPRM stub tube. The leakage is not isolable from the reactor vessel. The reactor was in Mode 4 at the time of discovery. "This event is being reported as a degraded condition pursuant to 10CFR50.72(b)(3)(ii)(A)." A repair plan is being formulated. The licensee has notified the NRC Resident Inspector.

They usually have two sumps in containment. One called identified leakage and the other called unidentified leakage and they measure the leakage on a hourly basis. The non identified leakage sump collects from mostly water from the non primary water cooling water systems like the service water system or component cooling water system. Say for the reactor coolant pump motor bearing cooling. A little leakage from non pressure barrier leakage is not risky and you can delay shutting down. The identified leakage can only come from pressure barrier leakage. 

Is something wrong with NRC inspector training? Why didn't they immediately order a plant shutdown?  

But the deal here is, you positively have to know it’s not coming from a primary system pressure barriers like the seal water system or potentially like leakage from the reactor vessel. The idea you have two pressure barrier leaks is chilling.

So far I haven't found any LERs on this kind of leakage. Is this kind of leak a industry first? It looks like that according to the NRC documents. Is this Susquehanna's Indian Point baffle bolt crack problem? You have to assume there is more flaws or leaks in the tube. Has the tubes ever been UT'd. What if the Uting of the tubes they discovered many more flaws and cracks throughout the tube. You have to assume all the tubes are degraded. You would have to UT all the tubes. I believe these guys just came out of a outage on May 4, 2016. Why didn't pressure testing of the primary system at end of outage detect the leakage. These leaks can't develop in a month. Did they falsify the testing paperwork?

This is going to be a prolonged outage for weeks and maybe months.  

LPRM Housing

I'd like to know what the pressure of the reactor vessel when the employees measured the leakage. The assumption at power is the water was at approximately 500 degrees and at least 500 psig. The leaking water immediately turned into steam. The steam turns into water when in contact with cooler components and maybe on the surface of the containment. Plus from the bottom of the air cooler in the containmen. This water will then drain into the non identified leakage sump. If the reactor is cooled down then the pressure is about zero. There is a big difference in measured leakage between a zero pressure plant and one at 500 psig and degrees. Besides the water leakage rate, it's an additional large artificial heat load into containment.  

I might got the identified and non identified leakage switched around. But it is the same concept.  

It will give you a false indication its not pressure barrier water. This is a well-known phenomenon. Like I said, employees can’t enter the containment at power because the radiation levels are too high. It is very difficult to tell whether its identified leakage or unidentified leakage. Regulations state, you have to be 120% sure you got non pressure barrier water leaking into containment or you must assume it is pressure barrier water. Then it is a mandatory emergency shutdown.

• Did they have or detect "containment" abnormal pressure or temperature symptoms indicating a high temperature leak?

• This is Davis Besse head leak. These plants monitor the containment air radiation levels. They sample the air radiation levels at least daily. Primary system water is very radioactive. So if pressure barrier water is leaking into the containment, it tremendously spikes up the air radiation level (particulate and gaseous). Davis Besse ignored the spiking containment air radiation levels. The air radiation level is really a sophisticated water leak detection system, it was put in the plant for that purpose. It amplifies the indications of detecting a small primary system leaks.        

If you got pressure barrier leakage, we can’t predict reliably how a crack would grow in a pipe or vessel. If it’s in the reactor vessel and the crack grows, it potentially overrides all of the safety designs of any plant. We are completely powerless to prevent a meltdown.  The NRC and licencees assume a reactor vessel crack and leak is a impossibility. So this accident is not designed into the plant. That is why it is really important to shutdown the plant immediately based on incomplete information.

General Technical Specification requirements and explanations 

PWR not BW: but same concept without steam generators. This is tech specs and constitutes a direct violation of regualtion.

B 3.4 REACTOR COOLANT SYSTEM (RCS) 

B 3.4.13 RCS Operational LEAKAGE BASES 

BACKGROUND 

Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE. 10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. 

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public. 

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leak tight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection. This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA). Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analyses for events resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is [one gallon per minute] or increases to [1 gallon per minute] as a result of accident induced conditions. The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than 150 gallons per day is significantly less than the conditions assumed in the safety analysis. 

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.
 RCS operational LEAKAGE shall be limited to: 

a. Pressure Boundary LEAKAGE 

No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. 

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid -is only briefly released via safety valves and the majority is steamed to the condenser. The [1 gpm] primary to secondary LEAKAGE assumption in the safety analysis is relatively inconsequential.
The [SLB] is more limiting for site radiation releases. The safety analysis for the [SLB] accident assumes [1 gpm] primary to secondary LEAKAGE in one generator as an initial condition. The dose consequences resulting from the [SLB] accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., a small fraction of these limits). 

The RCS operational LEAKAGE satisfies Criterion 2 of the NRC Policy Statement. Unidentified LEAKAGE (continued) 

One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary. 

c. Identified LEAKAGE 

Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system. 

d. Primary to Secondary LEAKAGE througqh Any One SG
The limit of 150 gallons per day per steam generator (SG) is based on the Operational LEAKAGE Performance Criterion in the Steam Generator Program. The Steam Generator Program criterion states: 

"The RCS operational primary-to-secondarv leakage through any one steam generator shall be limited to 150 gallons per day.' 

The RCS Operational primary to secondary LEAKAGE is measured at standard temperature and pressure.
The operational LEAKAGE rate limit applies to LEAKAGE in any one steam generator. If it is not practical to assign the LEAKAGE to an individual steam generator, all the LEAKAGE should be conservatively assumed to be from one steam generator.

This is really bad professionally. They made a dangerous assumption for a week or more the leak was non pressure barrier water. With the facts known today, they were immediately required to shut down.  The is not a hard call, if you know your limitations on differentiating non identified leakage from identified leakage, you just shut down the plant to fix the leakage.

This should be a red finding at least. They made a safety call on maliciously incorrect information.