Wednesday, February 14, 2018

LaSalle Torus Discovery Is A Example The NRC Doesn't Inforce Regulations

This is my example we don't know the true condition per licensing of every plant in the nation. We are going to have a lot of surprises in the next big accident... And the gap between licensing and the actual conditions of the plant are widening. Can you imagine all the processes though the decades that was missed by the licensee and NRC. It is horrible...  

LaSalle Inspection Report 

It is imperative a licensee knows the conditions of the safety equipment and all documentation reflects the actual conditions. The torus has probably been inop since the plant has been in operation. This isn't the case.

When found, the torus should have been declared inop and emediately shutdown till the paperwork has been fixed. Plus an additional amount of time. It would help keep everyone else keep safe fearing the NRC would thrown down the hammer on them.
(Closed) Unresolved Item 05000373; 05000374/2016001–01:  Adequacy of Changes to Pool Swell Analysis a. Inspection Scope During the 2016 first quarter integrated inspection period, the inspectors reviewed the operability evaluation associated with loss of coolant accident suppression pool analysis.  The inspectors identified an unresolved item involving changes to the methodology and design assumptions of the suppression pool analysis and whether those aforementioned changes provide a reasonable expectation that the affected systems, structures and components were operable. During the follow-up inspection activities to the Unresolved Item (URI), the inspectors reviewed LaSalle County Station, Units 1 and 2—Issuance of Amendments Re:  Request to Revise Suppression Pool Swell Design Analysis and the Facility Licensing Basis (CAC NOS. MF8702 AND MF8703); dated October 30, 2017.  The inspectors also reviewed Operability Evaluation OE 12–003; Potential to Increase Pool Swell Loads; Revision 5 and supporting calculations of record.  The inspectors determined the licensee’s operability evaluation provided a reasonable expectation of operability.  Based on this review, the inspectors sufficiently resolved these concerns and consider URI 05000373; 05000374/2016001–01 closed with no performance deficiencies identified; however, during this review, the inspectors identified one additional issue described below. This operability inspection constituted one sample as defined in IP 71111.15–05. b. Findings Primary Containment Structure, Suppression Pool Columns, Downcomer Vent and Downcomer Vent Bracing Did Not Meet Seismic Category I Requirements Introduction.  A finding of very low safety significance (Green) and an associated NCV  of 10 CFR Part 50, Appendix B, Criterion III, “Design Control,” was identified by the inspectors for the failure to ensure the adequacy of the design for the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing.  Specifically, the inspectors identified three representative examples where the licensee failed to perform adequate design calculations resulting in the design not being in conformance with Seismic Category I requirements as defined in UFSAR Sections 3.8.1.4.1, 3.8.1.5 and 3.8.6. Description.  UFSAR Table 3.2–1 delineated the primary containment structure and downcomer vent as Seismic Category I and meeting the quality assurance requirements of 10 CFR Part 50 Appendix B.  The suppression pool columns were part of the primary containment structure and support the drywell floor.  The columns were designed to transfer design loading from the drywell floor to basemat.

19
In UFSAR Section 3.8.1.1.1.1 described the primary containment as utilizing a Mark II over/under pressure-suppression configuration.  The primary containment consisted of a steel pressure vessel enclosed by a concrete shield wall both supported by a concrete basemat.  The primary containment was enclosed by the reactor building, a reinforced-concrete structure functioning as a secondary containment. The drywell was connected to the suppression chamber by downcomer pipes.  Steam that could be released in the drywell during a postulated loss-of-coolant accident was channeled through these downcomer pipes into the suppression pool where it is condensed thus effecting pressure-suppression.  This would result in a lower pressure and temperature. The downcomer vent pipes were braced at Elevation 697’-1” and Elevation 721’-0”.  The downcomer vent bracing design function was to provide horizontal restraint for applied lateral loading on downcomer vent pipes due to the seismic and loss-of-coolant accident design event.  The downcomer vent and downcomer vent bracing design requirements are delineated in Section 5.3.3.4 of LaSalle County Station, "Mark II-Design Assessment Report (LSCS-DAR)," Commonwealth Edison Company, Chicago, Illinois,  September, 1982.  The design assessment report was incorporated by reference in UFSAR Section 3.8.6. During a review of calculations for the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing, the inspectors identified the following three representative examples in which the licensee failed to meet the design requirements: • Calculation No. 195B; Containment Assessment; Revision 0; and Calculation  No. 161I; Suppression Pool Columns; Revision 0.  UFSAR Section 3.8.1.4.1 stated, in part, “The design and analysis procedure is in full compliance with the requirements of Article CC–3000 of the ASME B&PV Code, Section III, Division 2…” The design yield strength of reinforcement shall not exceed 60,000 psi as described in Section CC–3422 of Article CC–3000.  In addition, UFSAR Section 3.8.1.5 defined the allowable of Fy as the minimum guaranteed reinforcing steel yield strength.  The licensee used certified material test reports or actual material yield strength for the reinforcing steel in the evaluation of the containment structure and suppression pool columns.  The use of actual material yield strength did not meet American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Section III, Division 2 and UFSAR requirements.  The licensee documented these deficiencies in Issue Report No. 4070065; NRC Id:  Clarification on Material Strength Values in Calcs; dated October 16, 2017. • Calculation No. L–002547; Assessment of Containment Wall, Basemat, Liner, Reactor Pedestal, Downcomer Bracing, Drywell Floor, and Suppression Pool Columns for 105 percent Power Uprate; Revision 0.  As delineated in Section 5.3.3.4 of LaSalle County Station, "Mark II-Design Assessment Report, the stresses within the downcomer were considered acceptable if they satisfy the ASME B&PV Code, Section III, Subsection NE.  As permitted by Subsection NE–1120 for Metallic Containment components the downcomers were analyzed using Subsection  NB–3650 of Section III.  The licensee did not use the ASME code acceptance limits.  The licensee documented these deficiencies in Issue Report No. 4074674; NRC Id:  Clarification of Design Basis Code of Downcomer Vent; dated November 14, 2017.

20
• Calculation No. L–002547; Assessment of Containment Wall, Basemat, Liner, Reactor Pedestal, Downcomer Bracing, Drywell Floor, and Suppression Pool Columns for 105 percent Power Uprate; Revision 0.  Section 5.3.3.4 of LaSalle County Station Mark II-Design Assessment Report described the allowable acceptance limits are based on the 1.6 times the American Institute of Steel Construction (AISC) allowables but no greater than 0.95 times Fy (minimum specified yield strength of section).  The licensee increased the allowable stresses  by 50 percent based on using plastic section modulus properties which exceeded the elastic acceptance limits set forth in Section 5.3.3.4 of LaSalle County Station  Mark II-Design Assessment Report.  The use of plastic section modulus properties would allow for permanent deformation of the material.  Also, the downcomer bracing gusset plate uses plastic section modulus properties as well.  Lastly, the licensee used a dynamic increase factor of 10 percent to increase the allowable acceptance limits.  The dynamic increase factor was not contained in Section 5.3.3.4 of LaSalle County Station Mark II-Design Assessment Report.  The licensee documented these deficiencies in Issue Report No. 4070067; NRC Id:  Clarification on Acceptance Criteria in Calcs; dated October 16, 2017. The inspectors reviewed the operability evaluation in accordance with IMC 0326; Operability Determinations & Functionality Assessments for Conditions Adverse to Quality or Safety; dated November 20, 2017 to assess whether the nonconforming primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing were operable.  The inspectors identified no performance deficiencies with the operability evaluation.  In response to the inspector’s concern, the licensee initiated CAP documents as AR 4070067; NRC Id:  Clarification on Acceptance Criteria in Calcs; dated October 16, 2017, AR 4070065; NRC Id:  Clarification on Material Strength Values in Calcs; dated October 16, 2017 and AR 4074674; NRC Id:  Clarification of Design Basis Code of Downcomer Vent; dated November 14, 2017. Analysis.  The inspectors determined the licensee’s failure to perform adequate evaluations to demonstrate Seismic Category I compliance for the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing was contrary to the design control measures per 10 CFR Part 50, Appendix B, requirements and was a performance deficiency.   The performance deficiency was determined to be more than minor because the performance deficiency was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events.  Specifically, compliance with Seismic Category I design basis requirements was to ensure the primary containment structure, suppression pool columns, downcomer vents and downcomer vent bracing would function as required during a Seismic Category I design basis event and not adversely affect the function of the containment barrier.  

Friday, February 09, 2018

Am I Seeing More Of These kinds Of Pages?

Current Event Notification Report for February 9, 2018
U.S. Nuclear Regulatory Commission
Operations Center
Event Reports For
02/08/2018 - 02/09/2018
** EVENT NUMBERS **
***No events reported***




















Thursday, February 08, 2018

Grand Gulf: So unmazingly Unprofessional



February 7, 2018

Mr. Eric Larson, Site Vice President Entergy Operations, Inc. Grand Gulf Nuclear Station P.O. Box 756 Port Gibson, MS  39150

SUBJECT: GRAND GULF NUCLEAR STATION, UNIT 1, RESPONSE TO CLARIFICATION OF INITIAL RESPONSE TO A NOTICE OF VIOLATION AND NUCLEAR REGULATORY COMMISSION INSPECTION REPORT 05000416/2017012 

Dear Mr. Larson:

Thank you for your letter of December 21, 2017, (ML17362A041) which provided clarifying information to supplement your September 21, 2017, reply (ML17269A031) to the Notice of Violation concerning calibration of the main steam line and containment/drywell high range radiation monitors.  The Notice of Violation was issued on August 22, 2017, in NRC Inspection Report 05000416/2017012 (ML17235B265).  

The NRC identified several discrepancies in your initial response that required clarification.  We have reviewed these clarifications in conjunction with your initial response to the Notice of Violation and find that they address the violation and the corrective actions taken.  Following review of your responses to the Notice of Violation, we concluded that the surveillances performed on the main steam line and containment/drywell high range radiation monitors in response to the Notice of Violation demonstrate reasonably accurate instrument responses.  However, the changes to the calibration procedures themselves warrant additional review.  In particular, we noted that the procedures continue to allow tolerances significantly wider than industry norm.  We will review the implementation of your corrective actions during a future inspection to determine that full compliance has been achieved and will be maintained.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice and Procedure," a copy of this letter will be made available electronically for public inspection in the NRC Public Document Room or from the NRC’s Agencywide Documents Access Management System (ADAMS), accessible from the NRC’s Web site at http://www.nrc.gov/reading-rm/adams.html.    

Tuesday, February 06, 2018

The NEI, NextEra And Entergy War.

Is this the beginning of the great collapse with the USA's nuclear industry? Is the NEI playing hardball with NextEra trying to prevent other utilities from the leaving the organization?

NextEra Suit Accuses Nuclear Trade Group of ‘Extortion"

February 5, 2018

By Rich Heidorn Jr.

NextEra Energy, which quit the Nuclear Energy Institute last month over the trade association’s push for subsidies, last week accused the group of “extortion,” saying it was spitefully denying the company access to a database used to screen workers.


The company initially declined to say publicly why it was leaving NEI when it informed the organization of its decision on Jan. 4.
But NextEra ended its silence after NEI notified it on Jan. 30 that it was terminating its access to the Personnel Access Data System (PADS). NextEra said NEI informed it that it would be cut off Feb. 4 unless it paid $860,000, “the vast majority of which is NEI membership fees unrelated to PADS.”
“NEI’s actions were taken for no purpose other than to retaliate against the NextEra companies because of their withdrawal as NEI members,” said the suit, filed Feb. 2 in U.S. District Court for the Southern District of Florida.
NEI CEO Maria Korsnick issued a statement Monday saying she “vehemently denies” NextEra’s allegations and “will vigorously defend our position in court.”
NextEra said losing access to PADS could threaten seven scheduled refueling outages at its nuclear plants in 2018, including one set to begin Feb. 7 at the St. Lucie nuclear plant owned by its Florida Power & Light subsidiary. The company said St. Lucie’s workforce would jump from 700 to 1,700 during the monthlong outage.
The nuclear industry developed PADs in the mid-1990s as a shared database for employee security information such as criminal history reports, fitness-for-duty test results and psychological screenings.
NextEra said it would be “exceedingly difficult” to meet Nuclear Regulatory Commission requirements without PADS, noting that staff can more than double during plant outages. “Many of the additional maintenance workers employed during these refueling outages are highly transient — moving from plant to plant across the country to work during outages,” the company said. “Without access to PADS, nuclear operators would be forced to start from scratch in screening individual applicants for unescorted access, and they would do so without the benefit of consulting information already collected by other nuclear operators in an easily accessible electronic format. Similarly, without universal industry participation in PADS, the database would become incomplete. This would result in additional manual screening efforts even for continuing PADS participants.”
The company contends the PADS participation agreement, which it signed in 1995, does not require participants to be NEI members. “NEI took this retaliatory action notwithstanding that the NextEra companies have been at all times in compliance with the agreement and have paid millions of dollars to develop and upgrade PADS,” it said.

Korsnick disagreed with NextEra’s interpretation of the participation agreement. “When NextEra voluntarily chose to discontinue its NEI membership, it was no longer entitled to continue participating in PADS,” she said. “Even then, NEI conveyed to NextEra that it would supply the information in PADS necessary to maintain strict compliance with the NRC regulations. That exchange has been accomplished and will continue throughout each work week.
“To call NEI’s approach retaliatory, or even suggest the notion of extortion, is both counterfactual and offensive to the good faith effort the offer represents,” she continued. “NEI’s good faith outreach was intended to open a dialogue that would advance the industry’s interest in remaining unified, or as unified as possible, on regulatory and other policy positions. Unfortunately, rather than even opening a dialogue, NextEra immediately followed its rejection of NEI’s offer with a baseless lawsuit.”

Break over Policy

NextEra owns all or part of the Duane Arnold Energy Center in Palo, Iowa; the Point Beach Nuclear Plant in Two Rivers, Wisc.; and the Seabrook Station in Seabrook, N.H., equivalent to 6% of total U.S. nuclear generating capacity. In addition to the St. Lucie plant near Fort Pierce, Fla., FPL owns the Turkey Point plant near Miami. As of the end of 2016, NextEra also owned about 16% of U.S. wind capacity and 11% of the country’s solar capacity.
NextEra — which had been paying about $3 million in NEI dues annually — quit last month over what it called the trade group’s “irrational and unreasonable policies that would distort electric energy markets.”
Its suit cited NEI-funded studies “that call into question the reliability and costs of the electric system, attempting to create a false sense of panic and unfairly and incorrectly maligning the operations of its members, including the NextEra companies.”
“NEI claims that the ‘grid-based electricity supply portfolio in the United States is becoming less cost-effective, less reliable and less resilient,’” the complaint continues. “Such a thesis is unfounded. In fact, the policies that NEI is advocating would produce those very results by introducing artificial constraints on the way in which an electric system is planned and operated. … As large nuclear generators, the NextEra companies obviously support nuclear energy. But the NextEra companies cannot financially, or otherwise, support an organization that fundamentally mispresents the state of grid reliability in this country.”
Korsnick said NEI’s lobbying in support of Energy Secretary Rick Perry’s call for price supports for coal and nuclear plants followed “a rigorous process for gathering input from member companies to inform our policy positions.”
“On most issues [NEI] does not advocate a position until it has been approved by members of the Executive Committee. NextEra may not have agreed with NEI’s effort to support the continued operation of existing plants, but our work was guided by the interests of our member companies,” she said.
“NEI remains committed to achieving its foundational mission: to preserve, sustain, innovate and grow the nuclear energy industry. All of NEI’s actions should be and are consistent with that purpose. NEI also ensures all decisions and actions taken maintain a safe, effective and well operated nuclear energy fleet. NEI’s commitment to each of those core principles will always be absolute without compromise.”
NEI did not respond to a question about NextEra’s contention that the group is “suffering from financial difficulties.” NextEra cited NEI’s Form 990 for 2015, which it said “shows negative six-figure net assets for the 2015 and 2014 tax years.”

Entergy also Left NEI

Entergy, which operates seven nuclear plants in the U.S., also quit NEI last month, but it has not commented publicly on its reason for doing so.
“NEI has been one of several vehicles through which to advocate our positions on important policy and regulatory issues impacting the nuclear power industry,” Entergy spokeswoman Emily Bealke Parenteau said in response to a question about the company’s departure. “Entergy has made the decision to leverage its other internal and external resources for advocacy efforts.
“While Entergy will no longer be a member of NEI, we have a system in place that replaces PADS. We will continue to engage actively and cooperatively with the industry in both the operations and public policy arenas,” she added.
One industry official with knowledge of the situation said Exelon and some other NEI members view Entergy as a “traitor” for closing its uneconomic merchant nuclear plants rather than fighting for subsidies.
Exelon purchased Entergy’s James A. FitzPatrick nuclear plant in New York after the latter said it would close the plant regardless of whether the state approved zero-emission credits. Entergy also has agreed to close its Indian Point plant under pressure from Gov. Andrew Cuomo.
“Exelon told other NEI members that Entergy effectively forced them to buy [FitzPatrick] — they believed that … to get ZECs passed, they needed solidarity, and Entergy wasn’t playing ball,” the official said. “The fact that Entergy is closing Pilgrim [in Plymouth, Mass.] without a whimper and Palisades [in Michigan] when their contract ends in a few years has some NEI members upset. … Every time that a nuclear plant closes, it hurts their specialty vendors and, as a result, vendors shrink, and remaining ones have some market power. And that raises costs for every remaining plant.

Monday, February 05, 2018

New Startup For Junk Plant Grand Gulf

Jan 16
Grand Gulf 83%
River Bend 90%

Why has River Bend been stuck at 90% for 24 hours...

Update Jan 15
Hold your breath, Grand Gulf is near their peak power level (91%) last startup...

Grand Gulf  85%
River Bend 90%

Update Jan 14
Hmm, River Bend is pulling a Grand Gulf?

Grand Gulf  83%
River Bend 65%

Update Jan 13

Grand Gulf  63%
River Bend 85%

See this is driving me crazy. Why did Grand Gulf take a deep 22% drop in power, while River Bend is maintaining around 80% power? Why is Grand Gulf so unstable.

Update Feb 12

Jan 12 76% power

Jan 11 74%

Jan 10 55%

River Bend started up on Jan 10 (1% power) and is 81%

From the evidence, I just don't think it is a xenon problem. It might be some of it. I think something is going with the core or fuel.

Notice the event date and power level? The inner and outer airlock doors are supposed to be shut and tested before startup. It is not supposed to be opened until the next shutdown. On the far side of inner airlock door sits the reactor core. It is really radioactive inside that door. They had to open the door for a problem inside containment. I wonder if they actually went through the inner door a power? It is very abnormal. The failed seal in the inner door, they should have picked it up on the pre startup seal testing. I did leak rate testing on these doors all the time. They were inches thick metal door for employee radiation protection. The doors were very heavy. It took a lot of heft just to close the doors... I don't think a girl could close doors :)
Power Reactor Event Number: 53201
Facility: GRAND GULF
Region: 4 State: MS
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: BRANDON STARNES
HQ OPS Officer: DONALD NORWOOD
Notification Date: 02/10/2018
Notification Time: 22:37 [ET]
Event Date: 02/10/2018
Event Time: 18:35 [CST]
Last Update Date: 02/10/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(C) - POT UNCNTRL RAD REL
Person (Organization):
HEATHER GEPFORD (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 68 Power Operation 70 Power Operation

Event Text

BOTH INNER AND OUTER CONTAINMENT AIRLOCK DOORS INOPERABLE AT THE SAME TIME

"On 2/10/18 at 1835 CST at Grand Gulf Nuclear Station, while the 208 ft. Containment Airlock Outer Door was tagged-out for planned maintenance, the 208 ft. Containment Inner Door was determined to be inoperable. Grand Gulf had performed 06-ME-1M23-R-0001, Personnel Airlock Door Seal Air System Leak Test, on the 208 ft. Containment Airlock Inner Door which had been deemed satisfactory. While performing planned maintenance on the outer door an additional review of the paperwork determined that the test was actually unsatisfactory on the inner door. TS 3.6.1.2 Condition C was entered at 1835 CST on 2/10/18 for both 208 ft. Containment Airlock Doors being inoperable. Maintenance of the Outer Door is expected to be completed, and the airlock returned to operable status, prior to TS required action completion time."

The licensee notified the NRC Resident Inspector.


Feb 9

Must be a xenon soak thing. Up 3% to 70%.

River Bend is still 0% from last scram.

Feb 8

67%

Feb 7

Still at 26%. So we are already at a highly abnormal startup. They just weren't ready for power operation...

River Bend is still shutdown.

Feb 6

Stuck at 26% for 24 hours. The fun begins.

Feb 5 26%

Feb 4 17%

Feb 3 1%

Friday, February 02, 2018

Junk Plant Grand Gulf Another Scram: Homeless People Running the Plant

I made a mistake, My mind can't fathom the depth of what is going on at Entergy in the first swipe. So Grand Gulf has stayed shutdown. The power ascension of River Bend was ended with a scam.   

I know what they did wrong! They increased power to 100% too quickly (humor).

This is probably a instrumentation problem or employees inproperly calibrating the instrumentation. There is greater than a slight chance of reactor damage.


The Plant is in bad shape. There were two failure points here. Two things were broke or were mis-operated. What tripped the recirc pump and the flow indication problem. They made it a very fragile plant. Nobody really knows the extent of degraded and broken equipment at this facility.  

Power Reactor Event Number: 53192
Facility: RIVER BEND
Region: 4 State: LA
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: Timothy Gates
HQ OPS Officer: VINCE KLCO
Notification Date: 02/01/2018
Notification Time: 14:23 [ET]
Event Date: 02/01/2018
Event Time: 10:57 [CST]
Last Update Date: 02/01/2018
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
Person (Organization):
RICK DEESE (R4DO)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R Y 27 Power Operation 0 Hot Shutdown

Event Text

MANUAL REACTOR SCRAM

"At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3.

"The plant response to the scram was as expected. All control rods [fully] inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves.

"The NRC Senior Resident [Inspector] has been notified."