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{{Adams|number = ML070390148}}
{{Adams
| number = ML071000082
| issue date = 04/10/2007
| title = Clintion Power Station, Information Requested for NRC Biennial Baseline Component Design Bases Inspection 05000461/07-006
| author name = Stone A
| author affiliation = NRC/RGN-III/DNMS
| addressee name = Crane C
| addressee affiliation = Exelon Generation Co, LLC
| docket = 05000461
| license number = NPF-062
| contact person =
| document report number = IR-07-006
| document type = Letter
| page count = 12
}}


{{IR-Nav| site = 05000461 | year = 2007 | report number = 006 }}
{{IR-Nav| site = 05000461 | year = 2007 | report number = 006 }}


=Text=
=Text=
{{#Wiki_filter:
{{#Wiki_filter:ril 10, 2007
[[Issue date::February 7, 2007]]


EA-06-291Mr. Christopher M. CranePresident and Chief Nuclear Officer Exelon Nuclear Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555
==SUBJECT:==
CLINTON POWER STATION - INFORMATION REQUEST FOR AN NRC BIENNIAL BASELINE COMPONENT DESIGN BASES INSPECTION 05000461/2007006(DRS)


SUBJECT: FINAL SIGNIFICANCE DETERMINATION FOR A WHITE FINDING AND NOTICE OF VIOLATION; NRC INSPECTION REPORT NO. 05000461/2007006(DRS) FOR CLINTON POWER STATION
==Dear Mr. Crane:==
On October 15, 2007, the NRC will begin a biennial baseline Component Design Bases Inspection (CDBI) at the Clinton Power Station. This inspection will be performed in accordance with revised NRC Baseline Inspection Procedure (IP) 71111.21 and replaces the biennial Safety System Design and Performance Capability Inspection.
 
The CDBI inspection focuses on components which have high risk and low design margins.
 
The components to be reviewed during this baseline inspection will be identified during an in-office preparation week prior to the first week of the onsite inspection. In addition, a number of risk significant operator actions and operating experience issues, associated with the component samples, will also be selected for review.
 
The inspection will include three weeks onsite, including the information gathering during the first onsite week. The inspection team will consist of six NRC inspectors, of whom five will focus on engineering and one on operations. The current inspection schedule is as follows:
* October 15 through October 19, 2007;
* October 29 through November 2, 2007; and
* November 12 through November 16, 2007.
 
The team will be preparing for the inspection, mainly during the week of October 8, 2007, as discussed in the attached enclosure. A Region III Senior Reactor Analyst may accompany the inspection team during the week of October 15, 2007, to review probabilistic risk assessment data and assist in identifying risk significant components, which will be reviewed during the inspection.
 
Experience with previous baseline design inspections of similar depth and length has shown that these type of inspections are extremely resource intensive, both for the NRC inspectors and the licensee staff. In order to minimize the inspection impact on the site and to ensure a productive inspection for both parties, we have enclosed a request for information needed for the inspection. It is important that all of these documents are up to date and complete in order to minimize the number of additional documents requested during the preparation and/or the onsite portions of the inspection. Insofar as possible, this information should be provided electronically to the lead inspector. The information request has been divided into three groups:
* The first group lists information necessary for our initial inspection scoping activities.
 
This information should be available to the lead inspector no later than September 17, 2007. By September 28, the lead inspector will communicate the initial selected set of approximately 30 high risk components.


==Dear Mr. Crane:==
* The second group of documents requested are those items needed to support our in-office preparation activities. This set of documents, including the calculations associated with the initial selected components, should be available at the Regional Office no later than October 5, 2007. During the in-office preparation activities, the team may identify additional information needed to support the inspection.
The purpose of this letter is to provide you the final results of our significance determination of the preliminary Greater Than Green finding identified in Inspection Report No. 05000461/
2006011(DRS). The inspection finding was assessed using the Significance Determination Process and was preliminarily characterized as Greater Than Green, a finding of greater than very low safety significance, resulting in the need for further evaluation to determine significance; and therefore, the need for additional NRC action. This Greater Than Green finding involved the failure to select an appropriate method for calculating the minimum elevation (i.e., the analytical level) of water above the high pressure core spray (HPCS) pump suction line to preclude vortex formation and subsequent air entrainment in the pump's suction. As a result, the analytical level would result in significant air entrainment potentially causing the HPCS to be incapable of completing its safety function. At our request, a Regulatory Conference was held on December 19, 2006, to further discussyour views on this issue. The public meeting summary, including the handouts, can be found in the Agencywide Document Access and Management System (ADAMS) ML063520445. During the meeting, your staff described the results of recent scaled model testing. Specifically, the scaled model testing showed that a localized depression briefly formed which immediately collapsed and resulted in significant (about 24 percent) air being entrained in the suction piping. You determined that this air/water mixture would result in a slowed level decrease in the reactor core isolation cooling tank resulting in a delay in transferring suction to the suppression pool. However, your calculations showed that the suction valve from the suppression pool wouldopen and suction would be primarily from the suppression pool prior to the air reaching the suction of the HPCS pump. Therefore, you concluded that the HPCS pump would be capable of performing its safety function. In assessing the test model results, you assumed that flow C. Crane-2-would not be characterized as slug flow, that is, the flow would be less than 24 percent airentrained. Your conclusion was based on: (1) visual confirmation during Alden Laboratory Testing; (2) visual comparison of test results against a "known" flow having 24 percent air entrainment; and (3) a computer model (RELAP) prediction that slug flow would exist above 24 percent. The NRC identified the following concerns with your conclusion: *During testing at Alden Laboratory, the pump was stopped immediately (about 5 seconds) upon visual observation of 'break through', i.e., air becoming entrained in the suction pipe. This was done to preserve and prevent damage to the test pump.


This quick stopping did not allow time to verify the absence of slug flow.*The visual comparison of the test results against a "known" flow having 24 percent airentrainment was short in duration. This visual comparison may not represent actual test or in-plant flow conditions. With flow and flow conditions unstable and oscillating, this short time duration did not provide definitive proof that slug flow would not exist. Actual void fraction measurements may typically have as much as 8 percent uncertainty.*The computer model (RELAP) is a generic 2-phase flow code and is not necessarilytuned or calibrated to this exact scenario. To consider results 'exact' and without any consideration of analytical error is imprudent. The assumption of 24 percent air entrainment was key in assessing the ability of HPCS toperform its function for several reasons. First, the 24 percent provides a basis for the rate of decrease in the RCIC tank. A greater void fraction would slow down the rate of change, increasing the time to the swap-over point. This increase in time, would allow the air wave front to travel further down the line and potentially reach pump suction prior to full opening of the suppression pool suction valve. Secondly, although your staff calculated a 2-phase fluid flowvelocity, an increase in void fraction will increase the transport velocity, increasing the possibility of air arrival at the pump. Lastly, should slug flow exist, there is a potential for system waterhammer affecting system piping or the HPCS pump or both.In summary, the staff does not concur with your evaluation regarding the amount of airentrainment; and therefore, does not agree with your assessment on the past operability of the HPCS pump. Your assessment is not conclusive, complete or robust, in that the basis for 24 percent was not well founded. Small changes to these assumptions may significantly impact the conclusion regarding past HPCS pump operability. In addition, during the Regulatory Conference, you also provided your assessment of thesignificance of the finding. Specifically, you provided information regarding the potential for operators to throttle HPCS flow and the estimated contribution to the risk from fire events. The NRC reviewed the information regarding throttling the HPCS injection valve and determined C. Crane-3-that it should be considered in the final significance determination. Based on the discussion atthe Regulatory Conference, operators would be directed to throttle HPCS in response to transient (i.e., non- Loss of Coolant Accidents and non- Anticipated Transient Without a Scram)
* The last group includes the additional information above as well as plant specific reference material. This information should be available to the team on the first day of the inspection. It is also requested that corrective action documents and questions developed during the inspection be provided to the inspectors throughout the inspection.
scenarios. If operators successfully throttle the HPCS injection valve, the system flow rate will be low enough that air entrainment during suction swap-over to the suppression pool would no longer be a concern. For the final significance determination, the NRC assumed that HPCS would fail in response to transient initiating events only if the operator failed to properly throttle the HPCS injection valve. For all other initiating events, HPCS was assumed to fail during the suction transfer, consistent with the assumption in the preliminary significance determination. Given the inherent uncertainty in estimating human error probabilities, the NRC used its best estimate of 2.6E-2 for the human error probability in the final significance determination.The NRC also reviewed the estimation of fire risk contribution that you provided and determinedthat it was the best available information; and therefore, it was used directly in the final significance determination.After considering the information presented at the Regulatory Conference and the additionalinformation you provided in your letter dated December 21, 2006, the NRC has concluded that the inspection finding is appropriately characterized as White, an issue with low to moderate increased importance to safety, which may require additional NRC inspections. Using the estimation of fire risk contribution and best estimate for human error probability, the NRC determined the total change in core damage frequency to be about 4.4E-6 per year. You have 30 calendar days from the date of this letter to appeal the staff's determination ofsignificance for the identified White finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2.The NRC has also determined that the failure to ensure the adequacy of design of the HPCSsystem by performance of design reviews or by use of alternate or simplified calculational methods is a violation of Title 10 Part 50, Appendix B, Criteria III, as cited in the enclosedNotice of Violation (Notice). The circumstances surrounding the violation are described in detailin Inspection Report No. 05000461/2006011(DRS). In accordance with the NRC Enforcement Policy, NUREG-1600, the Notice of Violation is considered escalated enforcement action because it is associated with a White finding.You are required to respond to this letter and should follow the instructions specified in theenclosed Notice when preparing your response.Because plant performance for this issue has been determined to be in the regulatory responseband, we will use the NRC Action Matrix, to determine the most appropriate NRC response for this event. We will notify you, by separate correspondence, of that determination.


C. Crane-4-In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure and response will be made available electronically for public inspection in the NRC Public Document Room or from the Publically Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
The lead inspector for this inspection is Mr. Zelig Falevits. We understand that our licensing contact for this inspection is Mr. R. Frantz of your organization. If there are any questions about the inspection or the material requested in the enclosure, please contact the lead inspector at (630) 829-9717 or via e-mail at ZXF@nrc.gov.


Sincerely,/RA/James L. CaldwellRegional AdministratorDocket No. 50-461License No. NPF-62
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).


===Enclosure:===
Sincerely,
Notice of Violationcc w/encl:Site Vice President - Clinton Power StationPlant Manager - Clinton Power Station Regulatory Assurance Manager - Clinton Power Station Chief Operating Officer Senior Vice President - Nuclear Services Vice President - Operations Support Vice President - Licensing and Regulatory Affairs Manager Licensing - Clinton Power Station Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing Assistant Attorney General Illinois Emergency Management Agency State Liaison Officer, State of Illinois Chairman, Illinois Commerce Commission 1 HQ concurrence received via e-mail from D. Starkey, OE on February 2, 2007
/RA/
Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62 Enclosure: Component Design Bases Inspection Document Request See Attached Distribution
}}
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Latest revision as of 07:34, 23 November 2019

Clintion Power Station, Information Requested for NRC Biennial Baseline Component Design Bases Inspection 05000461/07-006
ML071000082
Person / Time
Site: Clinton Constellation icon.png
Issue date: 04/10/2007
From: Ann Marie Stone
Division of Nuclear Materials Safety III
To: Crane C
Exelon Generation Co
References
IR-07-006
Download: ML071000082 (12)
v • d • e


Text

ril 10, 2007

SUBJECT:

CLINTON POWER STATION - INFORMATION REQUEST FOR AN NRC BIENNIAL BASELINE COMPONENT DESIGN BASES INSPECTION 05000461/2007006(DRS)

Dear Mr. Crane:

On October 15, 2007, the NRC will begin a biennial baseline Component Design Bases Inspection (CDBI) at the Clinton Power Station. This inspection will be performed in accordance with revised NRC Baseline Inspection Procedure (IP) 71111.21 and replaces the biennial Safety System Design and Performance Capability Inspection.

The CDBI inspection focuses on components which have high risk and low design margins.

The components to be reviewed during this baseline inspection will be identified during an in-office preparation week prior to the first week of the onsite inspection. In addition, a number of risk significant operator actions and operating experience issues, associated with the component samples, will also be selected for review.

The inspection will include three weeks onsite, including the information gathering during the first onsite week. The inspection team will consist of six NRC inspectors, of whom five will focus on engineering and one on operations. The current inspection schedule is as follows:

  • October 15 through October 19, 2007;
  • October 29 through November 2, 2007; and
  • November 12 through November 16, 2007.

The team will be preparing for the inspection, mainly during the week of October 8, 2007, as discussed in the attached enclosure. A Region III Senior Reactor Analyst may accompany the inspection team during the week of October 15, 2007, to review probabilistic risk assessment data and assist in identifying risk significant components, which will be reviewed during the inspection.

Experience with previous baseline design inspections of similar depth and length has shown that these type of inspections are extremely resource intensive, both for the NRC inspectors and the licensee staff. In order to minimize the inspection impact on the site and to ensure a productive inspection for both parties, we have enclosed a request for information needed for the inspection. It is important that all of these documents are up to date and complete in order to minimize the number of additional documents requested during the preparation and/or the onsite portions of the inspection. Insofar as possible, this information should be provided electronically to the lead inspector. The information request has been divided into three groups:

  • The first group lists information necessary for our initial inspection scoping activities.

This information should be available to the lead inspector no later than September 17, 2007. By September 28, the lead inspector will communicate the initial selected set of approximately 30 high risk components.

  • The second group of documents requested are those items needed to support our in-office preparation activities. This set of documents, including the calculations associated with the initial selected components, should be available at the Regional Office no later than October 5, 2007. During the in-office preparation activities, the team may identify additional information needed to support the inspection.
  • The last group includes the additional information above as well as plant specific reference material. This information should be available to the team on the first day of the inspection. It is also requested that corrective action documents and questions developed during the inspection be provided to the inspectors throughout the inspection.

The lead inspector for this inspection is Mr. Zelig Falevits. We understand that our licensing contact for this inspection is Mr. R. Frantz of your organization. If there are any questions about the inspection or the material requested in the enclosure, please contact the lead inspector at (630) 829-9717 or via e-mail at ZXF@nrc.gov.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-461 License No. NPF-62 Enclosure: Component Design Bases Inspection Document Request See Attached Distribution

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