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.