ML20294A132

NRC-2016-000727 - Resp 3 - Final, Agency Records Subject to the Request Are Enclosed

ML20294A132
Person / Time
Site:	Saint Lucie
Issue date:	02/17/2017
From:	Stephanie Blaney NRC/OCIO/GEMSD/FLICB
To:	Lawrence Criscione - No Known Affiliation
Shared Package
ML20294A127	List: ML20294A128 ML20294A129 ML20294A130 ML20294A132 ML20294A133
References
FOIA, NRC-2016-000727
Download: ML20294A132 (67)

Text

Dear FOIA Requester:

The FOIA Improvement Act of 2016, which was enacted on June 30, 2016, made several changes to the Freedom of Information Act (FOIA). Federal agencies must revise their FOIA regulations to reflect those changes by December 27, 2016. In addition to revising our regulations, we intend to update the Form 464, which we use to respond to FOIA requests.

In the interim, please see the comment box in Part I.C of the attached Form 464. The comment box includes information related to the recent changes to FOIA that is applicable to your FOIA request, including an updated time period for filing an administrative appeal with the NRC.

Sincerely yours, S ~ B~ ISi Stephanie Blaney FOIA Officer

NRC FORM 464 Part I U.S. NUCLEAR REGULATORY COMMISSION FOIA RESPONSE NUMBER (12-2015) 1 2016-0727 3 RESPONSE TO FREEDOM OF 11 INFORMATION ACT (FOIA) REQUEST I I

RESPONSE

TYPE INTERIM FINAL REQUESTER: DATE:

!Lawrence Criscione 11 fEB1,-.

DESCRIPTION OF REQUESTED RECORDS:

Records corresponding to items 3 (ML16216A704) and 4 (ML16216A711 ), as further explained in the Comments Section, below.

PART I. -- INFORMATION RELEASED Agency records subject to the request are already available in public ADAMS or on microfiche in the NRC Public Document Room.

0 Agency records subject to the request are enclosed.

Records subject to the request that contain information originated by or of interest to another Federal agency have been referred to that agency (see comments section) for a disclosure determination and direct response to you.

We are continuing to process your request.

0 See Comments.

PART I.A - FEES AMOUNT*

D [ZJ

===-III You will be billed by NRC for the amount listed. None. Minimum fee threshold not met.

$ 11.!::=:I

• see Comments for details D You will receive a refund for the amount listed. D Fees waived.

PART I.B - INFORMATION NOT LOCATED OR WITHHELD FROM DISCLOSURE We did not locate any agency records responsive to your request. Note: Agencies may treat three discrete categories of law enforcement and national security records as not subject to the FOIA ("exclusions"). 5 U.S.C. 552(c). This is a standard notification given to all requesters; it should not be taken to mean that any excluded records do, or do not, exist.

D We have withheld certain information pursuant to the FOIA exemptions described, and for the reasons stated, in Part II.

Because this is an interim response to your request, you may not appeal at this time. We wilt notify you of your right to appeal any of the responses we have issued in response to your request when we issue our final determination.

You may appeal this final determination within 30 calendar days of the date of this response by sending a letter or email to the FOIA Officer, at U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001, or FOIA.Resource@nrc.gov.

Please be sure to include on your letter or email that it is a "FOIA Appeal."

PART I.C COMMENTS ( Use attached Comments continuation page if required)

In conformance with the FOIA Improvement Act of 2016, the NRC is informing you that you have the right to seek assistance from the NRC's FOIA Public Liaison.

In our interim responses dated December 22, 2016 and January 26, 20 I 7, we addressed items I (MLI 6236A0 19),

2 (MLl6236A021), 5 (MLl6238A013) and 6 (ML16238A0l4). This final response addresses the remaining two records requested in your request. Since the date of your request, all of these records have been removed from ADAMS.

[continued on next page]

Steyfianie ..'A. 'Bfane NRC Form 464 Part I (12-2015) Page 2 of3

NRC FORM 464 Part I U.S. NUCLEAR REGULATORY COMMISSION FOIA RESPONSE NUMBER (12-2015) 1.-----20-16--0-12_1__ II 3 RESPONSE TO FREEDOM OF INFORMATION ACT (FOIA) REQUEST Continued IZ]

RESPONSE

INTERIM FINAL TYPE REQUESTER: DATE:

!Lawrence Criscione 11 ,m 1 , 2011 PART I.C COMMENTS (Continued)

However, because the NRC was able to locate these records by the accession numbers when your request was received, we have processed the records (except as noted, below).

Item 3 (ML16216A704) appears to be a Linked In page of an individual when he was employed by NRC. Given its personal nature, and since NRC has not located other copies of this record in any of its record systems, we have determined that it is a personal record. As such, it is not subject to the FOIA and has not been processed.

Item 4 (ML16216A711) is a copy of a briefing package prepared by David Lochbaum of the Union of Concerned Scientists, ahead of meetings that were scheduled with Chairman Bums and Commissioner Baran, which the Commission confirmed was received. It is enclosed.

NRC Form 464 Part I (12-2015) Page 3 of 3

ucsu sa.org Two Brattle Square, Cambridge, MA 02138-3780 t 617.547.5552 f 617.864.9405 1825 I< Street NW, Suite 800, Washington, DC 20006-1232 t 202.223.6133 f 202.223.6162 2397 Shattuck Avenue, Suite 203, Berkeley, CA 94704-1567 t 510.843.1872 f 510.843.3785 One North LaSalle Street, Suite 1904, Chicago, IL 60602-4064 t 312.578.1750 { 312.578.1751 MARCH 13,, 2015 MATERIALS FOR MEETINGS WITH CHAIRMAN STEPHEN G. BURNS AND COMMISSIONER JEFF BARAN DAVID LOCHBAUM DIRECTOR, NUCLEAR SAFETY PROJECT

AGENDA

<D Lying to the American Public about Nuclear Safety

@ Improperly Withholding Information from the Public Lessons from Fort Calhoun

© UCS Annual Report on the NRC and Nuclear Plant Safety February 26, 20 15 Page 2

Lying to the American Public about Nuclear Safety

Background

On April 19, 2011, the NRC staff conducted the annual assessment meeting for the Oconee nuclear station in Seneca, South Carolina (ML l l l 1707829). The first of two bullets on slide 2 of the NRC staff's slideshow indicated that a purpose of the meeting was to provide:

• "A public.forum for discussion of the licensee's performance in 2010" With Slide 15, the NRC staff summarized a yellow and a white finding by NRC inspectors during 2010.

But at a public meeting conducted 5 weeks after flooding ,c aused three reactor meltdowns at Fukushima, the NRC staff failed to mention to the public that it had issued a Confirmatory Action Letter (MLI 2363A086) to Duke on June 22, 2010, requiring the company to take 15 measures to better protect the three reactors at Oconee from meltdown from flooding damage should the upriver Jocassee Dam fail.

The NRC staff had a tremendous opportunity to inform the public that, nine months prior to Fukushima, the NRC had identified similar flood protection vulnerabil ities at Oconee and had taken steps to ensure those vulnerabilities were addressed. In fact, several of the 15 measures had already been implemented while several others were far down the road to implementation.

But instead the NRC staff opted to play "duck and cover" and lie to the public.

The stated purpose of the meeting was to discuss licensee performance in 2010.

The licensee's performance in 2010 prompted the NRC to issue a Confirmatory Action Letter (CAL) in June 2010. CALs are rarely issued - the NRC staff issued more white findings in 2010 than CALs. The NRC staff chose to discuss its white finding at Oconee but remain silent about its CAL.

That incredibly poor judgment by the NRC staff undermined my trust and confidence in the agency. I now find it harder to believe it when the NRC staff says some condition is okay or that a problem has been resolved.

Given the staff's demonstrated propensity for hiding relevant information from the public and instead providing the public with a distorted, misleading version of nuclear plant safety, how can UCS and the public trust this agency to tell the whole truth and not just selective sub-truths?

February 26, 20 15 Page 3

Improperly Withholding Information from the Public

Background

In October 2004, the NRC staff sought and obtained Commission permission to withhold all incoming documents from licensees about fire protection and emergency planning (ML0423 l 0663). Since then. the NRC developed gu idance documents and revised regulations (10 CFR 2.390 in 2008) for licensees to ask NRC to withhold all or portions of documents they submit that contain sensitive security information. Despite this process being available for years, the NRC staff continues to withhold incoming fire protection and emergency planning documents, even when licensees do not request such withlholding.

Many of the withheld documents involved license amendment requests. By improperly withholding these documents, the NRC staff deprived the public of rights under federal regulations to contest requested actions.

The NRC staff has been handling submissions of Updated Safety Analysis Reports (USAR) oddly. Some USARs are placed into public ADAMS in their entirety (e.g., Beaver Valley Unit 2 at ML14339A408, Byron and Braidwood at ML1436A393, and Watts Bar Unit 2 at ML14155A256). Some USARs are withheld from public ADAMS in their entirety (e.g., Diablo Canyon per NRC memo at ML14022A120). The NRC staff has told the Senate EPW staff, the NRC OIG staff, and me three different stories last fa ll on why USARs may or may not be publicly available.

The USARs are key licensing documents, perhaps the single most important licensing document in existence. The USARs are heavily relied upon by licensees and NRC staff in preparing, reviewing, and approving operating license amendments. By improperly depriving the public of access to these vital documents, the NRC staff is unfairly impeding the public's ability to participate in licensing proceedings in a meaningful way.

That so many USARs are publicly available in ADAMS strongly suggests there is no legitimate reason for withholding the other USARs.

UCS and others frequently request NRC Communication Plans via the Freedom of Information Act. The NRC staff typically provides the requested plans with only personal privacy information (i.e., home telephone numbers) redacted (e.g., Salem/Hope Creek Safety Concious Work Environment issues at ML060620540, Oconee flood protection 50.54 letter at ML12326A389, Lndian Point CST pipe leak at MLI 10030931, Seabrook concrete degradation at MLl4161 A638, Davis-Besse concrete degradation at MLl41 7 1A27 1, etc.). But the NRC staff has also provided plans with all information, except page numbers, redacted contending the withheld information was "de liberative process" (Diablo Canyon seismic re-analysis at ML15033A280).

The NRC staff is playing games. The issues at Indian Point and Seabrook involved aging issues at a time when the reactors w ere seeking operating license renewals. The NRC staff provided essentially unredacted Communication Plans.

February 26, 201 5 Page 4

But the NRC staff redacted virtually the entire Communications Plan for Diablo Canyo n's seismic issues. True, the seismic issues are currently being monitored by the State and the NRC withjn an operating license renewal application proceeding, but again that was also the case at Indian Point and Seabrook.

UCS Recommendation UCS wrote to the NRC Chairman last November asking that the Commission reverse the policy of blanket withholding all incoming fire protection and emergency planning records.

UCS wrote to the NRC Inspector General asking that OIG investigate whether the agency violated federal regulations by approvi ng licensing requests about fire protection and emergency planning while denying the public access to the underlying documents.

The NRC should suspend issuing all operating licenses and approving all amendments to operating licenses until the agency has made publicly available all the documents it has been improperly withholding the past decade.

Withholding license amendment requests and USARs depiived the public its rights under federal regulations to participate in these licensing actions in a meaningful way. By improperly withholding these documents, the NRC staff is essenti ally giving its licensees uncontested proceedings and transforming purportedly open processes into closed, secret negotiations between the NRC staff and Licensees.

The NRC cannot contest the "cozy" label by bei ng "cozy" with licensees and denying the public its legal rights.

NOTE: UCS does not challenge the fact that certain information needs to be withheld. When information satisfies one or more of the criteria for withholding, then by all means withhold it.

But when information does not meet any of the criteria for withholding, then don' t withhold it.

NOTE: UCS also recognizes that given the sheer volume of documents handled by the NRC staff, there will be occasio nal mistakes made withholding some that should not be and di sclosing others that should be. UCS's concerns are not with the exceptions to the rule. UCS's concern is when the rule is mis-applied allowing many documents to be handled improperly.

February 26, 20 15 Page 5

Lessons from Fort Calhoun

Background

Fort Calhoun restarted in December 2013 following a 30-month outage to fix many longstanding safety problems.

It marked the 52nd time that a U.S. reactor remained shut down longer than a year to correct safety problems.

Fo11 Calhoun's outage began in Apri l 201 l, about a month after Fukushima.

The NRC formed a task force to extract lessons leamable from Fukushima and currently has a range of activities underway to implement those lessons.

The NRC did nothing to formally extract lessons learnable from Fort Calhoun.

Many of the safety problems that had to be fixed before NRC allowed Fort Calhoun to restart existed since 1996 or before.

Why had all the licensee's testing and NRC's inspections missed these safety problems?

Four times since the Reactor Oversight Process (ROP) was initiated, the NRC staff retuned Fort Calhoun to Action Matrix Column 1. Each time, the many safety problems that were finally fixed in 2011-20 13 had existed but were overlooked.

Twice since the ROP was initiated, the NRC staff returned Fort Calhoun to Action Matrix Column 2 from Column 3. Each time, the many safety problems that were finally fixed in 201 1-2013 had existed but were overlooked.

UCS Recommendation The NRC should formally evaluate Fort Calhoun's year-plus outage to identify lessons that enhance the effectiveness of its oversight efforts.

For example, the evaluation could take the safety issues on the NRC staff's Confirmatory Action Letter and reported to the NRC via Licensee Event Reports (LERs) from 2010 to 2014 and identify the NRC inspection procedures that examined these areas. These applicable inspection procedures could then be assessed to see whether changes in what gets examined or how it gets examined could have detected these problems. Similarl y, the evaluation might identify changes to the process used by the NRC staff to return Fort Calhoun to Action Matrix Columns l and 2 despite numerous safety problems that kept the reactor shut down for safety problems for 30 month. These might have been missed opportunities to have detected and corrected at least some of the many safety problems soone r.

Reference Document UCS Issue Brief "No More Fukushimas; No More Fort Calhouns," February 2015.

February 26, 201 5 Page 6

UCS Annual Report on the NRC and Nuclear Plant Safety

Background

UCS initiated a series of annual reports on the NRC and nuclear power plant safety in M arch 2011 . Each report summarizes the events the prior year that prompted the NRC to dispatch special inspection teams (SlTs) or augme nted inspection teams (AITs). Each report summarizes positive outcomes achieved by the NRC the prior year as well as negative outcomes.

This year's report noted that both the number and the severity of events triggering SITs/AITs continues a declining trend a nd acknowledges that NRC' s efforts very likely factored in these positive trends.

This year's report commends the NRC for undertaking two pro-active measures: the Reactor Oversight Process self-assessments and the Knowledge Management Program.

This year's report criticizes the NRC for improperly withholding documents from the public that denied meaningful participation in NRC's regulatory decision-making processes, for tolerating safety culture metrics that it found unacceptable when observed at nuclear plant sites and for subjecting two NRC engineers to recurring investigations because they voiced safety concerns.

UCS Recommendation The NRC instituted its Lessons Learned Program a decade ago. SECY-14-010 l (ML14175A780) is the most recent annual report on that program. It is a well-intended program gone terribly awry.

A total of merely seven ite ms were presented to the Lessons-Learned Oversight Board between August 2013 and May 2014. That list included only two reports from the NRC's Office of the Inspector General (OIG), no reports from the Government Accountability Office (GAO), none from the US Congress, and none from any external entity other than one classified, non-public DOE report.

It's virtually impossible to draw meaningful insights about trends and emerging problem areas from such paltry inputs. To be effective, the NRC's Lessons Learned Program must consider more inputs. For example, all OIG reports and GAO should be entered into the program.

Materials from external organizations should be reviewed for possible inclusion in the program.

The proliferation of inputs to the Lessons Learned Program would not require a linear increase in the fu ll-time equivalents needed to implement the program. The NRC staff responds to OIG and GAO reports. Thus, the additional work load for the Lessons Learned Program would be to monitor the findings and recommendations from the inputs seeking to identify common themes and whether a problem found here might also exist there.

Reference Documents UCS report dated March 20 15, "The NRC and Nuclear Power Plant Safety in 2014: Tarnished Gold Standard."

February 26, 201 5 Page 7

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 April 25, 2011 Mr. T. Preston Gillespie, Jr.

Site Vice President Duke Energy Carolinas, LLC Oconee Nuclear Station 7800 Rochester Highway Seneca, SC 29672

SUBJECT:

PUBLIC MEETING

SUMMARY

-OCONEE NUCLEAR STATION- DOCKET NOS. 50-269, 50-270 AND 50-287

Dear Mr. Gillespie:

This refers to the meeting conducted on April 19, 2011, in Seneca, SC. The purpose of this meeting was to discuss the NRC's Reactor Oversight Process (ROP) and the NRC's annual assessment of plant safety performance for the period of January 1, 2010, to December 31, 2010. The major topics addressed were the NRC's assessment program and the results of the assessment. A listing of meeting attendees and information presented during the meeting are enclosed.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter will be available electronically for public inspection in the NRG Public Document Room (PDR) or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS).

ADAMS is accessible from the NRG Web site at http://www.nrc.gov/readinq-rm/adams.html (the Public Electronic Reading Room).

Should you have any questions concerning this meeting, please contact me at (404) 997-4607.

Sincerely, IRA/

Jonathan H. Bartley, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos.: 50-269, 50-270, 50-287 License Nos.: DPR-38, DPR-47, DPR-55

Enclosures:

1. List of Attendees

2. Powerpoint Presentation cc w/encls: (See page 2)

Oconee Annual Public M eeting April 19, 201 l NAME AFFil,IATION

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3 Oconee Annual Public Meeting April 19, 2011 ,.,ce_

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2 rl Purpose of Today's Meeting

• A public forum for discussion of the licensee's performance in 2010

• Address the performance issues identified in the annual assessment letter 2 Protecting People and the Environment Enclosure 2

14 tl Oconee Assessment Results January 1 - December 31, 2010 Oconee Units 1, 2, and 3 were in the Degraded Cornerstone Column for all four quarters due to a Yellow Finding (Units 1, 2, and 3) and a White Finding (Units 2 and 3).

14 Protecting People and the Environment Enclosure 2

15 tl Safety Significant Findings or Pis

• Yellow Violation of TS 3.10.1 for SSF reactor coolant makeup subsystem inoperable for greater than allowed by technical specifications (Units 1, 2, and 3)

• White Violation of Criterion XVI, Corrective Action, for a failure to promptly identify and correct an adverse condition affecting operability of the Unit 2 and Unit 3 standby shutdown facility (Units 2 and 3) 15 Protecting People and the Environment Enclosure 2

UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 June 22, 2010 CAL 2-10-003 Mr. David A. Baxter Site Vice President Duke Energy Carolinas, LLC Oconee Nuclear Station 7800 Rochester Highway Seneca, SC 29672

SUBJECT:

CONFIRMATORY ACTION LETTER - OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3 COMMITMENTS TO ADDRESS EXTERNAL FLOODING CONCERNS (TAC NOS. ME3065, ME3066, AND ME3067) .

Dear Mr. Baxter:

This letter confirms commitments made by Duke Energy Carolinas, LLC (the licensee) in your June 3, 2010, letter. Specifically, the June 3, 2010, letter listed compensatory measures the licensee will implement at the Oconee Site and Jocassee Dam to mitigate potential external flooding hazards resulting from a potential failure of the JocasseeOam. The compensatory measures listed in the enclosure shall remain in place until final resolution of the inundation of the Oconee site from the failure of the Jocassee Dam has been determined by the licensee and agreed upon by the U.S. Nuclear Regulatory Commission (NRC), and all modifications are made to mitigate the inundation. The compensatory measures and implementation dates are set forth in the enclosure to this letter.

In addition to implementing the compensatory measures, pursuant to my telephone conversation with Mr. 13ill Pitesa of your company on June 22, 2010, you shall submit to the NRC by August 2, 2010, all documentation necessary to demonstrate to the NRC that the inundation of the Oconee site resulting from the failure of the Jocassee Dam has been bounded. Also, you shall submit by November 30, 2010, a list of all modifications necessary to adequately mitigate the inundation, and shall make all necessary modifications by November 30, 2011 .

Pursuant to Section 182 of the Atomic Energy Act, 42 U.S.C. 2232, you are required to:

1) Notify me immediately if your understanding differs from th~t set forth above;

2) Notify me if for any reason you cannot complete the actions within the specified schedule and advise me in writing of your modified schedule in advance of the change; and

3) Notify me in writing when you have completed the actions addressed in this Confirmatory Action Letter.

B-3

DEC 2 Issuance of this Confirmatory Action Letter does not preclude issuance of an Order formalizing the above commitments or requiring other actions on the part of the licensee; nor does it preclude the NRC from taking enforcement action for violations of NRC requirements that may have prompted the issuance of this letter. In addition, failure to take the actions addressed in this Confirmatory Action Letter may result in enforcement action.

This Confirmatory Action Letter will remain in effect until the NRC has concluded that all modifications necessary to adequately mitigate the inundation of the Oconee site from the failure of the Jocassee Dam has been completed.

Sincerely, IRA/

Luis A. Reyes Regional Administrator Docket Nos. 50-269, 50-270, 50-287 License Nos.: DPR-38, DPR-47, DPR-55

Enclosure:

Compensatory Measures cc w/encl: (See next page)

COMPENSATORY MEASURES NUMBER COMPENSATORY MEASURES IMPLEMENTATION STATUS 1 Perform flooding studies using the Hydrologic Engineering Complete Center -- River Complete Analysis System (HEC-RAS) model for comparison with previous OAMBRK models to more accurately represent anticipated flood heights in the west yard followina a postulated failure of the Jocassee Dam.

2 Maintain plans, procedures (Jocassee and Oconee) and Implemented

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guidance documents implemented (Oconee) to address 7

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w1 curren perspec~ves gameo owing the HEC-RAS sensitivity studies and the subsequent 2D inundation studies.

To the extent practical, the mitigation strategy is similar to existing extensive plant damage scenario (8.5.b) equipment, methods and criteria.

3 Duke Energy Hydro Generation will create a guidance Implemented document to consolidate river management and storm management processes. (Includes the Jocassee Development and the Keowee Development.)

4 Maintain a dam safety inspection program that includes: Implemented (1) weekly dam safety inspections of the Jocassee Dam by Duke Energy personnel, (2) dam safety inspections following any 2-inch or greater rainfall or felt seismic event, (3) annual dam safety inspections by Duke Energy, (4) annual dam safety inspections by FERC representatives, (5) five year safety inspections by FERC approved consultants. and (6) five year underwater inspections.

5 Maintain a monitoring program that includes: (1) continuous Implemented remote monitoring from the Hydro Central Operating Center in Charlotte, NC, (2) monthly monitoring of observation wells, (3) weekly monitoring of seepage monitoring points, and (4) annual surveys of displacement monuments.

6 Assign an Oconee engineer as Jocassee Dam contact to Implemented heighten awareness of Jocassee status.

• 7 Install ammeters and voltmeters on Keowee spillway gates Complete for eauipment condition monitorino.

8 Ensure forebay and tailrace level alarms are provided for Complete Jocassee to support timely detection of a developing dam failure.

9 Add a storage building adjacent to the Jocassee spillway to Complete house the backup spillway gate operating equipment (e.g.,

compressor and air wrench).

Enclosure

2 NUMBER COMPENSATORY MEASURES IMPLEMENTATION STATUS 10 Obtain and stage a portable generator and electric drive Complete motor near the Jocassee spillway gates to serve as a second set of backuo soillwav aate ooeratina eouioment.

11 Conduct Jocassee Dam failure Table Top Exercise with 06/30/2010 Oconee participation to exercise and improve response orocedures.

12 Instrument and alarm selected seepage monitoring locations 08/31/2010 for timelv detection of dearadina conditions.

13 Provide additional video monitoring of Jocassee Dam (e.g., 08/31/2010

. dam toe, abutments, and groin areas) for timely assessment of dearadina conditions.

14 Obtain and stage a second set of equipment (including a 11/30/2010 B.5.b-type pump) for implementation of the external flood mitiaation auidance.

15 Conduct Jocassee Dam/Oconee Emergency Response 12/31/2010 Organization Drill to exercise and improve response orocedures.

NOTES:

1. The word "complete" is used in the status column if the commitment regards a specific one-time equipment-related or analysis-related action that has been completed.

2. The word "implemented" is used in the status column if the commitment describes an on-going action that has been implemented.

Enclosure

POLICY ISSUE (Notation Vote)

October 19, 2004 SECY-04-0191 FOR: The Commissioners FROM: Luis A. Reyes Executive Director for Operations /RN

SUBJECT:

WITHHOLDING SENSITIVE UNCLASSIFIED INFORMATION CONCERNING NUCLEAR POWER REACTORS FROM PUBLIC DISCLOSURE PURPOSE:

To obtain Commission approval of guidance to be issued to the Nuclear Regulatory Commission (NRC) staff, power reactor licensees, and other agency stakeholders for withholding sensitive unclassified (nonsafeguards) information from public disclosure.

SUMMARY

In a staff requirements memorandum dated May 7, 2004, the Commission directed the NRC staff to develop guidance to ensure information that could reasonably be expected to be useful to potential adversaries is withheld from public disclosure. In determining whether information should be withheld or released , the NRC staff must attempt to appropriately balance our desire to maintain the openness of NRC's regulatory processes. with the need to protect the public from possible terrorist threats. This paper provides for Commission review and approval the NRC staff's proposed approach for determining the appropriate handling of information and more specific guidance for withholding or releasing information about nuclear power reactors (Attachment 1).

CONTACTS: William D. Reckley, NRR/IRT 301 -415-1323 Margie Kotzalas, NRR/IRT 301-415-2737

Subject Discussion and/or typical controls Test Program (Initial and lnservice Uncontrolled Inspections and Testing)

Accident Analysis Uncontrolled - Accident analyses typically included in licensing-related correspondence involve conservative models to demonstrate a plant's ability to respond to design basis transients (i.e., nonsecurity related events),

and is not treated as sensitive.

Technical Specifications (including Uncontrolled Bases)

Quality Assurance Uncontrolled Fire Protection Incoming documents are initially profiled as nonpublic -

staff will review for release upon request. Most information related to fire protection will not need to be designated as sensitive. Drawings showing details such as the specific location of equipment, doorways, stairways, etc. are to be withheld under 10 CFR 2.390.

Emergency Planning Incoming documents are initially profiled as nonpublic -

staff will review for release upon request. Most information related to emergency planning will not need to be designated as sensitive. Special attention is needed to determine if information relates to the response by a licensee or government agency to a terrorist attack. Note that some State and local governments consider parts of their emergency plans to be sensitive.

Security Information related to security programs at nuclear reactors is generally designated as SGI and is protected in a manner similar to classified confidential information.

Security-related information within the inspection program and reactor oversight process is withheld from public disclosure under 10 CFR 2.390.

Risk-Informed Decisionmaking Uncontrolled - exceptions include information related to (e.g., documents related to risk- security activities (e.g ., vulnerability assessments) and informed licensing actions, information related to uncorrected configurations or accident sequence precursor conditions that could be useful to an adversary. Special (ASP) analyses, significance attention should be applied to this area and information determination process (SOP) should be withheld if it describes a vulnerability or plant-notebooks, design certifications) specific weakness that is more helpful to an adversary than are the insights provided in open source literature .

Detailed computer models have been and will continue to be withheld from public disclosure.

• I *

• Beaver Valley Power Station FENOC' P.O. Box 4 Shippingport.PA 15077 RrstEnergy Nuclear Cperating Ccmpany Eric A. ~arson. 724-682-5234 Site Vice President Fax: 724-643-8069 November 24, 2014 L-14-360 10 CFR 50.71(e) 10 CFR 50.54(a) 10 CFR 54.37(b)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-001

SUBJECT:

Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Submittal of the Updated Final Safety Analysis Report, Revision 21 In accordance with the requirements of 10 CFR 50.71(e), the FirstEnergy Nuclear Operating Company (FENOC) is hereby submitting to the Nuclear Regulatory Commission (NRC) the Beaver Valley Power Station (BVPS), Unit No. 2, Updated Final Safety Analysis Report (UFSAR) Revision 21 in CD-ROM format. This submittal reflects facility and procedure changes implemented between November 2, 2012 (the end of Refueling Outage 16), and May 23. 2014 (the end of Refueling Outage 17), along with several changes implemented after Refueling Outage 17.

In accordance with NRC guidance for electrof!ic submissions, Attachment 1 provides a listing of the document components that comprise the enclosed CD-ROM. In addition to the UFSAR, the CD-ROM includes the BVPS, Unit No. 2 Licensing Requirements Manual, Revision 81 , and the Technical Specification Bases, Revision 27. The Technical Specification Bases are submitted in accordance with Technical Specification 5.5.10.d, "Technical Specifications (TS) Bases Control Program."

In accordance with 10 CFR 50.54(a), FENOC is hereby submitting a copy of the current revision of the FENOC Quality Assurance Program Manual (QAPM). The QAPM, Revision 19, is included in the enclosed CD-ROM.

Attachment 2 includes a summary of information removed from the BVPS, Unit No. 2 UFSAR in accordance with Appendix A to Nuclear Energy Institute (NEI) 98-03, "Guidelines for Updating Final Safety Analysis Reports," Revision 1.

Beaver Valley Power Station, Unit No. 2 L-14-360 Page2 FENOC conducted a review of BVPS, Unit No. 2 plant changes for 10 CFR 54.37(b) applicability. No components were determined to meet the criteria for newly identified components as clarified by Regulatory Issue Summary (RIS} 2007-16, Revision 1, "Implementation of the Requirements of 10 CFR 54.34(b) for Holders of Renewed Licenses."

There are no regulatory commitment changes to be submitted in accordance with NEI 99-04, "Guidelines for Managing NRC Commitment Changes."

This certifies, to the best of my judgment and belief, that Revision 21 of the BVPS, Unit No. 2 UFSAR accurately presents changes made since the previous submittal that are necessary to reflect information and analysis submitted to the Commission or pursuant to Commission requirements.

This letter contains no new regulatory commitments. If you have any questions regarding this report, please contact Mr. Thomas A. Lentz, Manager - Fleet Licensing, at 330-315-6810.

Sincerely, Eric A. Larson Attachments:

1. Document Components on CD-ROM

2. Information Removed from the BVPS, Unit No. 2 UFSAR

Enclosures:

Beaver Valley Power Station, Unit No. 2 UFSAR, Licensing Requirements Manual, Technical Specification Bases, and QAPM (on CD-ROM) cc: NRC Region I Administrator NRC Resident Inspector NRC Project Manager Director BRP/DEP (without Enclosures)

Site BRP/DEP Representative (without Enclosures)

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.:pw bpo!AOAM&>AMS Exelon Generation Byron/Braidwood Nuclear Stations Updated Final Safety Analysis Report (UFSAR)

Revision 15 December 2014 Byron Station, Units 1 and 2 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF""66 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-454, STN 5-0-455, and 72""68 NRC Docket Nos. STN 50-456, STN 50-457, and 72-73

I IRrsflwor3f>7~87ur-0-eBa~a~sE'1~ dtt¥>1l@~~bl i c ADAMS Attachment 1 to be withheld from Public Disclosure Under 10 CFR 2.390. When separated from this Enclosure, this letter is decontrolled.

[IE Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 May 30, 2014 10 CFR 50.4 10 CFR 50.34(b) 10 CFR 2.390(d)(1)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington , D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Docket No. 50-391

Subject:

WATTS BAR NUCLEAR PLANT (WBN)-UNIT 2-FINAL SAFETY ANALYSIS REPORT (FSAR), AMENDMENT 112

References:

1. TVA letter to NRC dated February 13, 2014, "Watts Bar Nuclear Plant (WBN) - Unit 2 - Final Safety Analysis Report (FSAR), Amendment 111 "

2. TVA letter to NRC dated May 8, 2014, "Watts Bar Nuclear Plant (WBN)

Unit 2 - lnservice Test (1ST) Program/Preservice Test (PST) Program" This letter transmits WBN Unit 2 FSAR Amendment 112 (A 112), which reflects changes made since the issuance of Amendment 111 on February 13, 2014 (Reference 1). contains a summary listing of FSAR sections and corresponding Unit 2 change package numbers associated with the A 112 FSAR changes.

FSAR A112 is contained on the enclosed Optical Storage Media (OSM #1 ) (Attachment 1).

The FSAR contains security-related information identified by the designation "Security-Related Information -Withhold Under 10 CFR 2.390." TVA hereby requests this information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390. A redacted version of the FSAR is contained on OSM #2 (Attachment 2),

which is suitable for public disclosure. contains a listing of the FSAR pages that have been redacted. Enclosure 3 lists the files and file sizes on the security-related OSM (OSM #1 ), and Enclosure 4 lists the files and file sizes on the publicly available OSM (OSM #2).

In regard to Supplemental Safety Evaluation Report (SSER), Appendix HH Open Items, the following can be stated to address three open items:

U.S. Nuclear Regulatory Commission Page 2 May 30, 2014 For Open Item No. 1, involving power assisted cable pulls. WBN Unit 2 construction has not made nor will not be making any such power assisted cable pulls in the completion of WBN Unit 2 . A 112 addresses Open Item No. 35, involving Component Cooling System (CCS), and Open Item No. 91, involving Feedwater Purity.

In addition, FSAR Change Package 2-112-10 addresses a clarification to the 1ST Program code of record as committed to in Reference 2.

Attachment 3 provides replacement disks for Amendment 111 provided in Reference 1.

During the course of Amendment 112 preparation, it was discovered that the discs containing the Amendment 111 files previously provided by Reference 1 did not contain Section 6 .2 .6.

Enclosures 5 and 6 have been updated to reflect this addition for file sizes related to the security-related and the publicly available OSMs for Amendment 111 .

There are no new commitments made in this letter. This letter does not close any "Generic Communications." If you have any questions. please contact Gordon Arent at (423) 365-2004.

I declare under the penalty of perjury that the foregoing is true and correct. Executed on the 30th day of May, 2014.

~*c;;/;y{)_ .

Raymond A. Hruby, Jr.

General Manager, Technical Services Watts Bar Unit 2

Enclosures:

1. WBN Unit 2 FSAR A 112, "Summary Listing of A 112 FSAR Changes"

2. WBN Unit 2 FSAR A 112,

• summary of Redacted Pages"

3. WBN Unit 2 FSAR A 112, "List of files and file sizes on the security-related OSM (OSM #1)"

4. WBN Unit 2 FSAR A 112. "List of files and file sizes on the publicly available OSM (OSM #2}"

Attachments:

1. OSM #1 : WBN Unit 2 FSAR Amendruent 112 - Security-Related Information - Withhold Under 10 CFR 2.390

2. OSM #2: WBN Unit 2 FSAR Amendment 112 - Publicly Available Version

3. OSM #1 : WBN Unit 2 FSAR Amendment 111 - Security-Related Information - Withhold Under 10 CFR 2 .390 OSM #2: WBN Unit 2 FSAR Amendment 111 - Publicly Available Version cc: See Page 3

June 23, 2014 MEMORANDUM TO: Michael T. Markley, Chief Plant Licensing IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation FROM: Peter J. Bamford, Project Manager IRA/

Plant Licensing IV-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

SUBJECT:

DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - REVIEW OF FINAL SAFETY ANALYSIS REPORT UPDATE, REVISION 21 (TAC NOS. MF2945 AND MF2946)

This memorandum documents the in-office review of Revision 21 to the Final Safety Analysis Report (FSAR) Update for Diablo Canyon Power Plant (DCPP), Units 1 and 2, dated September 16, 2013 (not publicly available). The FSAR Update was submitted by Pacific Gas and Electric Company (PG&E, the licensee), in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.71 (e). PG&E follows the guidance of Nuclear Energy Institute (NEI) 98-03, Revision 1, "Guidelines for Updating Final Safety Analysis Reports," and NEI 99-04, Revision 0, "Guidelines for Managing NRC Commitment Changes."

The time requirements for FSAR submittals are stated in 10 CFR 50.71(e)(4). Revisions must be filed annually or 6 months after each refueling outage provided the interval between successive updates does not exceed 24 months. In its letter dated December 8, 1997, the licensee requested an exemption from the time requirements stated in 10 CFR 50.71(e)(4) for DCPP, Units 1 and 2. As discussed in the licensee's exemption request, DCPP, Units 1 and 2, have a common FSAR. The rule would require FSAR updates within 6 months of each refueling outage, resulting in required FSAR updates every 12 months. As such, the licensee requested an exemption to allow the updates of the FSAR to be submitted within 6 months after each DCPP, Unit 2, refueling outage, but not to exceed 24 months from the last update. The Nuclear Regulatory Commission (NRC) staff approved the exemption in a letter dated March 12, 1998 (ADAMS Accession No. ML022400141 ). DCPP, Unit 2, completed its last refueling outage on March 23, 2013. The previous update of the DCPP FSAR, Revision 20, was submitted on November 16, 2011 (ADAMS Accession No. ML11332A181 ). Therefore, the September 16, 2013, submittal date for Revision 21 of the DCPP FSAR meets the requirements approved in the exemption since the submittal was within 6 months of the last DCPP, Unit 2, refueling outage and does not exceed 24 months from the last FSAR update.

As stated in the licensee's letter dated September 16, 2013, Revision 21 of the DCPP FSAR contains changes to reflect the plant configuration as of March 23, 2013. This meets the requirement in 10 CFR 50.71 (e)(4) which states that the revisions must reflect all changes up to a maximum of 6 months prior to the date of filing.

M. Markley Amendments Revision 21 covered changes to the FSAR Update during the period June 6, 2011 , through September 16, 2013. Each of the license amendments issued during the period were reviewed for impacts on the FSAR Update and included Amendment Nos. 211/213 through 216/218 (for Units 1 and 2, respectively). The following three amendments were identified which resulted in impacts on the FSAR Update:

• Amendment Nos. 211/213, dated March 29, 2012 (ADAMS Accession No. ML120790338), modified FSAR Update Sections 8.1.4.3, "Regulatory Guides," and 8.3.1.1.13.1, "Diesel Generator Unit Description," to identify an exception to Revision O of Regulatory Guide 1.9, "Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants";

• Amendment Nos. 212/214, dated October 31, 2012 (ADAMS Accession No. ML120300114), modified FSAR Update Sections 15.2.7.3, "Results," and 15.2.16, "References," to adopt a new analysis methodology for establishing the reduced power range neutron flux high setpoint for one inoperable main steam safety valve; and

• Amendment Nos. 214/216, dated January 9, 2013 (ADAMS Accession No. ML12345A379), modified FSAR Update Section 4 .3.2.2, "Power Distribution," to allow the use of the Best Estimate Analyzer for the Core Operations-Nuclear (BEACON) Power Distribution Monitoring System methodology, as described in Westinghouse Electric Company LLC's WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operation Support System," January 2000.

The FSAR Update changes for Amendment Nos. 211/213 were not apparent in Revision 21 . The licensee had reorganized the FSAR Update, removing the numbered Sections 8.1.4.3 and 8.3.1.1.13.1. However, the licensee included the amendment's language in Section 8.3.1.1.6.3.13, "Safety Guide 9, March 1971 - Selection of Diesel Generator Set Capacity for Standby Power Supplies," and Section 8.3.1.1.6.1. 13, "Safety Guide 9, March 1971 - Selection of Diesel Generator Set Capacity for Standby Power Supplies." With the inclusion of this exception in these two sections, the NRC staff concludes that the FSAR Update is consistent with the updates stated in Amendment Nos. 211/213.

Inspection Reports The inspection reports (IR) for the appropriate period were reviewed. The first, IR 2012004, involved a non-cited violation of Appendix B, Criteria V, "Instructions, Procedures, and Drawings," after PG&E failed to promptly evaluate the operability of plant structures, systems, and components (SSCs) after a newly discovered local fault line. The IR, dated February 14, 2012 (ADAMS Accession No. ML120450843), indicated a need to update the FSAR Update with the new seismic information. The second, IR 2011005, dated November 13, 2012 (ADAMS Accession No. ML12318A385), involved a Severity Level IV violation where the licensee failed to update the FSAR Update with information describing how plant SSCs meet 10 CFR Part 50,

ML13155A238), documented an event in which the licensee identified an unanalyzed condition due to a nonconservative change in the FSAR Update Chapter 15, "Accident Analyses," which would have resulted in a higher received radiological dose received by control room operators d uring an accident, but would not exceed General Design Criteria 19. The LER described the corrective actions taken to address the event and NRC staff confirmed that Revision 21 of the FSAR Update incorporated the corrective actions described in the LER.

The NRC staffs sampling review of the FSAR Update, Revision 21 included the applicable amendments, I Rs, and LERs. The staff did not find any commitments to modify the FSAR Update in its review. Based on the review, the staff conclludes that the FSAR Update, Revision 21 was submitted consistent with the requirements in 10 CFR 50.71 (e).

Docket Nos. 50-275 and 50-323 DISTRIBUTION:

PUBLIC LPL4-1 R/F RidsNrrDorllpl4-1 Resource RidsNrrLAJBurkhardt Resource RidsNrrPMDiabloCanyon Resource RidsRgn4MailCenter Resource ADAMS Accession No. ML14022A120 OFFICE NRR/DORL/LPL4*2/PM NRR/DORL/LPL4* 1/PM NRR/DORL/LPL4- 1/LA NRR/DORL/LPL4* 1/BC NRR/DORL/LPL4*1/PM NAME MOrenak PBamford JBurkhardt MMarkley PBamford DATE 6/17/1 4 6/17/1 4 6/17/14 6/23/14 6/23/14

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'61 I ICIAE use OICE I 81!UU:mlT I l<ELA I ED ,UfiiE?PPQJ,Ott COMMUNICATIONS PLAN Davis-Besse Nuclear Power Plant Steam Generators Replacement Inspection January 2014 Point Of

Contact:

Atif Shaikh, RIii 630-829-9824 GOALS

• Be prepared to answer public questions on the steam generators replacement inspection

• Be prepared to answer internal questions on the steam generators replacement inspection KEY MESSAGES The NRC's oversight of the steam generator replacement process at Davis-Besse is comprehensive to ensure the safety of the plant and the public.

Inspections started on December 2, 2013, and these inspections will continue through the actual replacement installation work beginning in February 2014 the post installation tests performed by the licensee, and the plant's subsequent return to power. The results of this NRC inspection will be documented in a publically available report that will be issued by the NRC within 45 days of the conclusion of this inspection.

NRC inspectors will conduct direct observations along with reviews of records, calculations, and procedures to provide adequate assurance that the plant modifications associated with the replacement steam generators meet applicable regulatory requirements.

Inspections will be conducted by a team of inspectors with expertise in metallurgy, structural design, heavy loads, radiatiol) protection, security, and other relevant areas.

NRC inspectors will review the licensee's evaluation of relevant steam generator replacements operating experience (OpEx) to determine whether the licensee has adequately evaluated the OpEx potentially relevant to the Davis-Besse steam generators repla~ement.

NRC inspectors will ensure that any safety concerns identified during the inspection are adequately addressed by the licensee.

The NRC staff invited the public to listen in via conference call to its initial inspection planning meeting with the licensee during which the licensee provided a presentation and NRC staff answered questions from the public. That presentation remains available to the public in the NRC's ADAMS document system (ML No. 13078A249) via the NRC public web site.

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NRC staff also discussed inspection plans with the public during the last end-of-cycle meeting near the plant and provided information in a meeting with local government officials. In addition, the NRC staff also plans to conduct a webinar to answer questions from the public related to the replacement steam generators at Davis-Besse.

BACKGROUND Davis-Besse is a Babcock and Wilcox (B&W) designed plant. It is a two loop plant and has two steam generators. The original steam generators are B&W designed once-through steam generators (OTSGs). The new replacement steam generators are also B&W designed OTSGs.

There are two basic types of steam generators used in the United States: recirculating steam generators (RSGs) and OTSGs. RSGs have tubes that are shaped like an inverted "U" while OTSGs have straight tubes. There are currently 59 units in the U.S. with RSGs and 6 units with OTSGs.

All steam generators are designed to limit the possibility of tube-to-tube contact since such a condition can result in the tubes rubbing against each other and leading to tube thinning. The thinning of the tube wall due to the interaction of two structures (e.g., tube-to-tube or tube-to-support) is commonly referred to as tube wear.

In Early 2012 , the licensee for San Onofre Nuclear Generating Station Unit 3, which has recirculating steam generators, detected hundreds of tubes with wear attributed to tube-to-tube contact caused by a fluid-elastic instability. Some of these indications were significant including one that leaked during normal operation and led to the plant shutting down. These indications occurred after approximately 20 months of operation. In total, eight tubes were found that did not meet the structural integrity performance criteria specified in the plant's technical specifications. The steam generators at San Onofre were designed and fabricated by Mitsubishi Heavy Industries (M HI).

In early 2010 , Three Mile Island, Unit 1 (TM l-1), completed the replacement of both its original OTSGs with new OTSGs that were fabricated by AREVA (France). The first inservice inspection of the TMl- 1 replacement steam generators took place in fall 2011 . During these inspections at TMl- 1, the licensee detected several tubes with indications. A more detailed in_vestigation led the licensee to conclude that these indications were a result of tube wear due to tube-to-tube contact.

In fall of 2013 the licensee for TMl- 1 conducted their second inservice inspection of the replacement steam generators. The licensee reviewed their testing data and concluded that tube-to-tube wear was progressing slowly "as predicted" based on first cycle wear data from fall of 2011.

In spring 2006, Oconee, Unit 3 conducted the first inservice inspection of the replacement OTSGs that were installed in 2004. The inservice inspection results revealed widespread wear degradation of the tubing at tube support plant (TSP) locations. Oconee, Units 1 and 2, have also experienced this widespread tube wear degradation at TSP locations following the first cycle of operation since installation in 2004. In spring of 2012 the licensee for Oconee, Unit 3 also detected wear attributed to tube-to-tube contact in the replacement OTSGs. The Oconee replacement OTSGs were designed and fabricated by B&W Canada and are similar to the design of the Davis-Besse replacement OTSGs.

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J iFFl'ilO I 1155 Obi! X a SFGPRIIX PS: OTil!I ::c: CIC.Iii tlllU The licensees for Oconee and TMI evaluated the severity of the tube-to-tube wear indications in their replacement steam generators. These evaluations concluded that the wear indications did not compromise tube integrity (i.e., the tubes could still perform their intended function consistent with their original design and licensing basis). In addition, this tube-to-tube contact did not involve high energy fluid-elastic instability such as that experienced at SONGS. NRC staff reviewed the licensees' evaluations and did not identify any safety issues that would affect plant restart.

Q&As FOR DAVIS-BESSE STEAM GENERATORS REPLACEMENT

1. Will this be a like for like replacement?

No, this will not be a like for like replacement. Although the replacement steam generators (SGs) are manufactured by the same vendor as the original SGs, there are some differences in the design of these replacement SGs. Hence, the licensee is required to perform an evaluation consistent with Section 50.59 of Title 10 to the Code of Federal Regulations ( 10 CFR) for the proposed modifications associated with the replacement SGs.1

2. What are the differences between the old and new steam generators?

The differences between the original SGs and the replacement SGs a/I relate to physical design aspects such as the material, component dimensions, number of tubes per generator, etc. The required design and safety functions of the SG remain the same.

The NRC staff will be reviewing the 50. 59 analyses supporting the design changes to ensure that plant safety is not impacted by the changes and to evaluate licensee's conclusions regarding whether NRC approval is needed for the changes.

3. Can you explain the 50.59 process?

The 50.59 process involves implementation of the requirements set forth in 10 CFR 50.59, a federal regulation . Essentially, whenever a licensee decides to implement a physical change to its facility or change how the facility is operated, used or controlled, including changes to safety analyses or documentation (e.g., a calculation , evaluation, methodology), then the 50.59 regulation allows a licensee to implement that change without prior NRC approval only if the change meets criteria pertaining to the safety implications of the proposed change. Generally, if a change would place the plant outside of the safety boundaries established by the NRC and reflected in the plant's licensing basis (e.g., NRC regulations, licensing documents, and plant safety analyses report), then prior NRC approval would be needed.

4. Can you explain the license amendment process?.

In general, the license amendment application revi~w process has 5 steps: 1)

Conducting an acceptance review to determine if there is sufficient technical information for the NRC staff to begin a detailed technical review of the application; 2) Publishing a Federal Register notice that describes the application and gives members of the public an opportunity to comment on the proposed determination of No Significant Hazards Consideration (NSHC) and request permission to be a party in a hearing; 3) Conducting a technical review to determine the safety of, and the environmental impacts of, the proposed amendment, including, if needed, sending requests for additional information (RAls) to obtain additional information needed to make an informed regulatory decision;

4) Completing the NRC staff's safety evaluation (SE), which provides the technical, OFSICI Oh: rtl8E 8Ub>.E 8206td I I "LL:ltll!D llkOAMSlt 11014 3

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Of POI* Is I lGli 9111slf &liiURrJ,c Rlils ft iliQ IIIEOPM0 IIOtl safety, and legal basis for the NRC's decision on the amendment application: and 5) If the amendment is approved, issuing the amendment and publishing a Federal Register notice that indicates when the amendment issued and whether the NRC staff made a final NSHC determination.

5. How do 50.59 analyses and license amendments assure safety?

Both processes provide assurance that changes at operating reactors are not made until the safety significance of the change is considered. As noted above, the 50.59 process can lead to a determination that a 50.90 license amendment application. and thus prior NRC approv*a1. is required .

6. What changes would require a license amendment?

If a proposed change is not consistent with a technical specification or places the plant outside of the safety boundaries established in the plant's licensing basis, then the change would require a license amendment.

7. Why not require a license amendment for the whole replacement?

NRC inspectors review samples of licensee 50.59 evaluations and decisions during the SG replacement inspections. If the Agency determines that a license amendment is required, the Agency can take appropriate enforcement action.

8. Are any license amendments needed for the SG replacements at Davis-Besse?

Davis-Besse submitted a license amendment request for Technical Specifications (TS) changes related to the replacement steam generators. *The NRC staff is currently reviewing this amendment request.

9 . Have any concerns been raised regarding the steam generator replacement?

A request for hearing and petition to intervene on the Technic~I Specification (TS) license amendment request was filed in May 2013. The petitioners challenged the 10 CFR 50.59 analyses on the steam generators replacement. contending that the steam generator replacement activities required an additional license amendment request. On August 12, 2013, the Atomic Safety Licensing Board (ASLB) denied the petition, The ASLB ruled that petitioners cannot challenge 10 CFR 50.59 analyses done to support steam generator replacement activities in a proceeding on a license amendment request to change TS related to operation with the new steam generators replacement. The ASLB also ruled that a ~hallenge to adequacy of 10 CFR 50.59 analyses for replacement of the steam generators can only be made by filing a petition under 10 CFR 2.206. *

10. Will the NRC staff conduct an inspection concerning the steam generator replacement activities?

Yes. The NRC staff will inspect the licensee's SG replacement activities during inspections which began on December 2, 2013. During the inspection , the NRC staff will review10 CFR 50.59 analyses done to support the steam generator replacement, as well as monitor steam generator replacement activities. An inspection report will be issued to document the results of the NRC staff's review.

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11. Will the NRC's review of the new steam generators/50.59 evaluations be complete before the plant can start up with the new steam generators?

It is the licensee's responsibility to ensure changes associated with the new steam generators are thoroughly evaluated and are safe and implemented appropriately . While the NRC staff will complete its inspection review as expeditiously as possible, we can't guarantee we will reach final conclusions prior to plant restart. The NRC staff will take the time it needs to do a thorough and rigorous inspection and to arrive at supportable conclusions. However, if at any time the NRC staff concludes that the changes are not safe, the NRC would take appropriate enforcement action, including ensuring the plant stays in or is placed in a safe condition.

12. Will there be an NRC inspection report for the DB steam generators? Will the inspection results be publicly available before restart?

The inspection results for the SG replacement inspection will be documented in a publicly available NRC inspection report which will be issued within 45 days after the completion of the inspection. The NRC inspection is extensive and includes evaluation of licensee activities that occur throughout the replacement outage and subsequent startup. Hence, the inspection report will not be available prior to startup.

13. Has the NRC incorporated lessons learned from previous SG replacements in inspections for the Davis-Besse replacements?

Recent operating experience at facilities where SGs have been replaced is being incorporated (or was incorporated) into the inspection effort for the Davis-Besse SG replacements. Region Ill staff closely coordinates with NRC headquarters to identify areas for a rigorous review of 50.59 evaluations. For the Davis-Besse steam generator replacement inspection, the NRC will be reviewing the licensees' evaluation of previous operating experience, key design differences between original and replacement steam generators, and if they exist, design change challenges discussed between the licensee and its vendor.

14. Has Davis-Besse licensee reviewed the SONGS or other SG replacement operating experience such as at TMl-1 and Oconee Unit 3 in preparation for their steam generator replacements?

Yes, Davis-Besse described in a public meeting how they have considered the SONGS, TMI , and Oconee SG tube degradation operating experience in their steam generator design and replacement activities. The NRC inspectors will review this information and the 50.59 evaluations supporting these design modifications as part of the SG replacement inspection activities.

15. Are these new steam generators considered an experimental design?

No, these new replacement SGs are not considered an experimental design. They are similar in basic design to the original SGs. There is also operating experience available regarding replacemeAt steam generators of a similar design as those being installed at Davis-Besse. The NRC inspectors will be reviewing the licensee's evaluation of the operating experience available as it pertains to the specific design.

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16. What are the main differences between the steam generators at Davis-Besse and SONGS?

• The Davis-Besse and SONGS SGs are different designs.

The stea m generators at SONGS are recirculating steam generator design. They are designed for a Combustion Engineering plant which requires larger steam

• generators, averaging close to 9,000 tubes per steam generator. The SONGS SGs were manufactured by MHI and are one of the largest steam generators used in the industry. The SONGS replacement SGs were modeled for vibration using MHl's proprietary modeling code.

• The Davis-Besse Steam generators are a completely different design from SONGS in that they are once through steam generators (they do not have a U-bend tube region, instead they consist of straight tubes) and were manufactured by B&W Canada. The Davis-Besse replacement SGs were modeled for vibration using an industry accepted EPRl modeling code.

17. Will DB cut a hole in the shield building for these replacement steam generators?

What impact will that cutting and opening process have on the existing shield building cracking?

In order to remove the old steam generators and install the new steam generators, the licensee will cut another hole in the reinforced concrete shield building. The hole will be located entirely within the boundaries of a previous hole that was cut for replacement of the reactor pressure vessel closure head, and hence will be in new concrete that was poured in 2012. Thus, the licensee does not expect there to be any impact on previously identified cracking in the older portions of the shield building wall.

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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION IV 1500 E. LAMA~ BLVD.

ARLINGTON. TX 76011~511 September 11, 2014 MEMORANDUM TO: Wayne Walker, Chief Division of Reactor Projects, Branch A FROM: Multiple Addressees. as listed below

SUBJECT:

COMMUNICATIONS PLAN - DIABLO CANYON POWER PLANT TOPICS OF INTEREST The purpose of this memo Is to transmit and request comments/concurrence on the enclosed Communications Plan for Diablo Canyon Power Plant (DCPP). The enclosed document is based on several iterations of informal communication plans, Q&A documents. and responses to congressional questions developed primarily by Region IV. NRR. OPA. and OCA over the last several years.

This communication plan describes the methods and resources that NRC staff will use to communicate with internal and external stakeholders regarding the DCPP seismic history and ongoing seismic evaluations being conducted in response to the Japan Lessons Learned Near-Term Task Force recommendations. Additionally, as applicable to current issues of interest to DCPP stakeholders, this communications plan integrates key messages related to spent fuel/dry cask storage and waste confidence issues (primarily by referencing other active .

communication plans).

This revision also incorporates Q&As for the most recent issues of concern including the licensee's AB-1632 Report to the State of California and the *sewell Report."

Once finallzed, the Communications Plan will be posted on the OEDO Communications website for use by the communications team and more broadly across the agency as necessary.

Most of those on concurrence have each provided significant input to iterations of this document (or documents from which this Plan was developed). As such, we are requesting your review/comments/concurrence In the next few daya (due by COB, Monday, September 16). Please forward your comments/concurrence on the document to Theresa Buchanan CTheresa.Buchanan@orc.gov and/or ph: (817) 200-1503) of my staff.

The concurrence block noted on the next page will be used to document your concurrence on the enclosed Communications Plan.

Enclosure:

As stated

OFFICIAL USE ONLY -- SENSITIVE INTERNAL INFORMATION (b)(5)

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Page 65 OFFICIAL USE ONLY SENSITIYE INTERNAL. INFORMATION

Paul Gunter Jim Ricc io Tirn Judson Dave Lochhaum Lucas H ixson Beyond Nuclear Greenpeace N uclear Information and Union of Concerned www.Enformable.com Resource Service Sc ientists November 19, 2014 Dr. AJlison M. Macfarlane, Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Dear ChaiJman Macfarlane:

On behalf of the Freedom of Information Team, I respectfully ask the Commission to revisit and revise the information withholding policies approved in Staff Requirements Memorandum (SRM) dated November 9, 2004, for SECY-04-0191 dated October 19, 2004.

In response to the tragic events of 9/1 1, the NRC staff proposed a framework for withholding information from the public that might be useful to adversaries attempting radiological sabotage at NRC-licensed facilities. The Commission approved the staff's proposal. In tlhe second paragraph of the SRM, the Commission directed that "the staff should move expeditiously to complete the necessary determinations and restore public access to the appropriate documents."

Since that time, the NRC and the nuclear industry have developed a system for withholding the proper information. For example, the NRC released Regulatory Issue Summary RIS-05-026, "Control of Sensitive Unclassified Nonsafeguards Information Related to Nuclear Power Reactors;" RIS-05-03 1, "Control of Security-Related Sensitive Unclassified Non-Safeguards Informatio n Handled by Individuals, Firms, and Entities Subject to NRC Regulation of the Use of Source, Byproduct, and Special Nuclear Material;" RIS-07-04, "Personally Identifiable Information Submitted to the U.S. Nuclear Regulatory Commission;" and RIS-12-03, "Reintegration of Security into the Reactor Oversight Program Assessment Program." The NRC also revised IOCFR 2.390 to clarify what information must be withheld.

The nuclear industry and the NRC have operating experience using this system. Today, there is a common understanding of what information needs to be withheld along with the appropriate means for withholding it.

It is now time to restore public access to the appropriate documents while retaining necessary protection against inappropriate disclosures.

Specifically, we ask that the framework in Attachment J to SECY-04-0191 profiling all incoming documents from plant owners about fire protection and emergency planning as nonpublic be reversed. All incoming documents about fi re protection and emergency planning should be profiled as public.

Plant owners now have clarity from the NRC regarding the nature and context of information that must be withheld from the public. Plant owners now also have an established and well-used process for submitting documents containing such information to the NRC so that the information is appropriately withheld. Thus, documents about fire protection or e mergency planning containing sensitive information will be submitted by plant owners per 10 CFR 2.390 and collateral processes, obviating the need for blanket withholding of all fire protection and emergency planning documents.

We look forward to the NRC restoring public access to appropriate fire protection and emergency planning information.

Sincerely, David Lochbaum Director, Nuclear Safety Project Union of Concerned Scientists PO Box 15316 Chattanooga, TN 37415 423-468-9272, office dlochbaum @ucsusa.org November 19, 2014 Page 2

ucsus.a.org T wo Bratt le Square, Cambridge, MA 02138-3780 t 617.547.5552 f 617.864.9405 1825 K Street NW, Suite 800, Wash ington, DC 20006-1232 t 202.223.6133 f 202.223.6162 2397 Shattuck Avenue, Suite 203, Berkeley, CA 94704-1567 t 510.843.1872 f 510.843.3785 One North LaSalle St reet , Suite 1904, Chicngo, IL 60602-4064 t 312.578.1750 f 312.578.1751 December 17, 2014 Hubert Bell , Inspector General U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Dear Mr. Bell:

On behalf of the Union of Concerned Scientists, I respectfully ask the Office of the Inspector General to investigate whether the Nuclear Regulatory Commission violated federal statutes and/or federa l regulations with the information withholding policy approved in Staff Requirements Memorandum (SRM) dated November 9, 2004, for SECY-04-0 191 dated October 19, 2004.

Among other things, the policy authorized the NRC staff to withhold all documents it received from plant owners involving fire protection and emergency planning. In the text on page 7 of the attachment to SECY 0 I 9 1, the NRC staff recognized that most of these incoming fire protection and emergency planning records would not likely contain sensitive information that needed withholding from the public. Yet the NRC staff recommended, and a majority of the Commission approved, withholding these incoming records.

Earlier this year, I submitted requests under the Freedom of Information Act for fire protection and emergency planning records dated October 1, 2004, or later that were not already publicly available. The fire protection records provided to me in response to my FOIA requests are mostly contained in the October 3, 2014, folder in the NRC's Agencywide Document Access and Management System (ADAMS).

No documents were withheld in their e ntirety by the NRC when responding to my FOIA requests. And I have not yet located a single redaction in any of the fire protection records 1

released by the NRC staff in response to my FOIA requests. Thus, there was no justifiable basis for withholding these records from the public.

1 Some of the emergency planning records released in response to my FOIA requests had telephone numbers and similar information redacted, but those redactions represented considerably less than one percent of the material in the documents.

Prinred on 100% posr-consumer recycled paper

But even if the tragic events of 9/ l l warranted error on the side of caution, a policy decision cannot trump or negate federal statutes and regulations. This policy with regard to fire protection and emergency planning records seems to have authorized practices that violate federal statutes a nd regulations. Several examples that strongly suggest that NRC violated federal statutes and regulations are summarized in the followi ng table.

Table 1: Some of the Fire Protection Records Withheld by the NRC Date ADAMS Document Document Made Comment ML Date Public The NRC approved the exemption Response to NRC request for on 09/27/2006. The approval additional information (ML062160387) was made public ML060300439 regarding fire suppression Ol/13/2006 10/03/2014 on 10/02/2006. Lack of access to exemption rnquest at Turkey the exemption request prevented or Point Units 3 and 4 sig nificantly impaired the public's ability to oppose it.

The NRC approved the exemption Response to NRC request for on 09/27/2006. The approval additional information (ML062160387) was made public ML0620 10J40 regarding fire suppression 07/12/2006 10/03/2014 on 10/02/2006. Lack of access to exemption request at Turkey the exemption requesl prevented or Point Units 3 and 4 sig nificantly impaired the public's ability to oppose it.

The NRC issued the amendment on 04/25/2007. The amendment (ML071 I 6043 I) was made public License amendment request on 05/17/2007. Notice of the for fire protection pending amendment was published ML063200100 11/15/2006 10/03/2014 requirements at Browns Ferry in the Federal Register on Units 1, 2, and 3 04/05/2007. Lack of access to the amendment request prevented or sig nificantly impaired the public's ability to oooose it.

The NRC issued license amendments on 09/16/2009. The amendment (ML082280465) was Supplement to license made public on 09/24/2008. Notice amendment request for of the pending amendments was ML082590007 deviation from fire protection 09/05/2008 10/03/2014 published in the Federal Register on requirements at South Texas 08/25/2009. Lack of access to the Project Units I and 2 deviation request prevented or sig nificantly impaired the public's ability to oppose it The NRC issued license amendments on 09/16/2009. The Response to NRC request for amendment (ML082280465) was additional information made public on 09/24/2008. Notice regarding requested deviation of the pending amendments was ML093350537 11/20/2009 10/03/2014 from fire protection published in the Federal Register on regulations at South Texas 08/25/2009. Lack of access to the Project Units I and 2 deviation request preven1ed or significantly impaired the public's ability to oppose it December J 7, 20 14 Page 2

Table 1: Some of the Fire Protection Records Withheld by the NRC Date ADAMS Document Document Made Comment ML Date Public The NRC approved the exemption on 03/11/2010. The approval Request for exemption from (MLI 00340670) was made public ML090570050 fire protection regulations at 02/18/2009 10/03/20 14 on 03/12/2010. L,Lck of access to FitzPatrick the exemption request prevented or significantly impaired the public' s ability to oooose it The NRC approved the exemption Response to NRC request for on 03/11/20 LO. The approval additional information (ML100340670) was made public ML090960214 regarding fi re protection 03/30/2009 10/03/20 14 on 03/12/2010. Lack of access to regulation exempt.io n request. the exemption request prevented or at FitzPatrick significantly impaired the public's ability to oppose it Licensee event report (LER) While LERs do not constitute for deficiencies in Appendix R licensing action requests (e.g.,

ML091320440 05/1 1/2009 10/03/2014 fire respo nse plan at Poi nt license amendments, exemptions, Beach Unit 1 deviations, etc.), they describe violations of regula tory requirements, either hardware or process related. When available, LERs could be cited by the public Licensee event report (LER) in opposing licensing requests for non-compliance manual involving hardware and process ML103570032 12/22/2010 10/03/20 14 actions in fire response plans changes. By withholding all fire at Monticello protection LERs, the NRC significantly hampered the public's ability to evaluate fire protection program adequacy and contest perceived shortcomings.

The NRC prepared its fi nding of no significant hazards for the Federal Register on 02/25/2010. The notice (ML! 00560391) was made public License amendment request to on 03/15/201 0. The NRC issued the ML093641067 use fire-resistive electrical 12/16/2009 10/03/2014 amendment on 09/30/2010. The cable at Wolf Creek amendment (ML I 02560498) was made public on 10/01/2010. Lack of access to the amendment request prevented or significantly impaired the public's ability to oppose it.

By withholding license amendment requests, the NRC seems to have violated 10 CPR

50. 9 1, Notice for public conunent; State consultation. Even when the agency publishes notices about the requests in the Federal Register, withholding the underlying request rendered that opportunity for public comment meaningless. The public lacked viable means to contest "secret" requests.

December J7, 20 14 Page 3

10 CPR 50.91 also provides opportunities for States to review proposed licensing actions and comment on or oppose them. The NRC's information withholding policy may also have infringed on States' abilities to conduct their consultation function. We request that OIG's investigation also determi ne whether the NRC' s policy adversely affected the States' role in licensing actions.

The NRC's information withholding policy would also seem to violate the spirit if not the letter of the Administrative Procedure Act. This federal statute requires agencies like the NRC to provide for public participation in rulemaking processes. While the fire protection and emergency planning records withheld by the NRC may not directly involve rulemaking, there most certainly is an indirect nexus. When plant owners requested exemptions from NRC's regulations promulgated via a public rulemaking process, the NRC deprived the public of its right to contest how the APA-compliant requirements were applied to the licensed nuclear facilities in their communities. And when the NRC pursued rulemaking, as it is and will be doing regarding emergency planning in response to both Fukushima's lessons and numerous reactor decommissionings, the NRC's withholding of the past decade's worth of emergency planning records essentially turned the APA-compliance rulemakings into a mockery of meaningful public participation. An oft-cited adage states that "information is power." The NRC's information withholding practice rendered the public powerless to participate in the agency's rulemaking proceedings.

A long with several other NGO representatives, I met with the NRC staff about document classification and information redaction policies on October 7, 2014, in a public meeting attended by some members of the OIG staff. We followed up with a letter to Chairman Macfarlane dated November 19, 2014, requesting the Commission to reverse the policy for withholding all incoming records involving fire protection and emergency planning. We have reason to bel ief the information withholding policy will be changed in the near future.

While we are hopeful that the NRC staff will soon cease blanket withholding of incoming fire protection and emergency planning records, that will solve only part of the problem.

We respectfully request that OIG investigate the policy to address the remainder of the problem. Even if the information withholding policy was justifiable, policy cannot violate federal statutes and regulations. Thus, the policy adopted by the NRC in late 2004 should not have resulted in requests for license amendments, deviations, and exemptions of fire protection regulatory requirements being withheld from the public.

December J7, 20 14 Pagc4

The information withholding policy adopted by the NRC in late 2004 attempted to better protect the public's safety. In applying the policy, the NRC undemli.ned the public's rights.

Thus, the NRC's good intentions were offset by the unintended consequences. The OIG 's investigation would identify those consequences as well as factors that could have or should have enabled maximum benefits to be derived with nli.nimal consequences. The report on the OIG's investigation can help the NRC staff implement process fixes that better maintain the delicate balance between the legitimate need to withhold some information and the public's right to know the rest of the information.

Sincerely, David Lochbaum D irector, Nuclear Safety Project Union of Concerned Scientists PO Box 15316 Chattanooga, TN 37415 423-468-9272, office dlochbaum @ucsusa.org December 17, 20 14 Page 5

[ C~'i!.ce~ned Scientists ISSUE BRIEF No More Fukushimas; No More Fort Calhouns HIGHLIGHTS Two significant nuclear power safety events occurred in the spring of 2011.

On April 9, 2011, operators shut On March 11, an earthquake and the tsunami it spawned caused the meltdown of down tlw 1w1ctor at the Fort Calhoun three reactors at the Fukushima Daiichi nuclear plant in Japan. Less than a month later, on April 9, operators shut down the reactor at the Fort Calhoun nuclear nuclear plant in Nebraska for a routiue plant in Nebraska for a routine refueling outage. But myriad safety problems dis-refueling outage. But myriad safety covered during the outage- many dating back to when the plant was constructed prob/em.1 discowred during the outage- in the late 1960s and early 1970s-prevented the reactor from restarting for two ma11y dati11g back to when the plant was and a half years.

constructed in the latr 1960s and early Following the first event, the U.S. Nuclear Regulatory Commission (NRC),

1970s-prevented the reactor from which oversees the safety of the nation's nuclear power plants, formed a task force that examined the Fukushima accident and identified more than 30 lessons that

,*estartingfor two a11d a halfyears. The could reduce vulnerabilities in the United States. The NRC ordered plant owners U.S. Nuclear Regulatory Commi.~sion to implement specific safety upgrades and is pursuing additional measures to (NRC), which oversees the 11atio11's nuclear further reduce vulnerabilities.

power pla11ts, needs to determine how its Following the second event, the NRC made no such effort to examine the Fort i11spectors and the plant ow11er missed-or Calhoun situation. lt failed to identify lessons that would enable it to detect safety dismissed rwmcrous longstanding safet.v violations sooner and correct them before they could accumulate to epidemic problems for years despite thousands of proportions requiring years to fix-or worse, contribute to an American Fukushima.

Fort Calhoun received its first operating license in 1973, and the NRC reli-hours ofi11spections. l t should appoi11t a censed the plant in 2003 to continue operating for as long as 20 more years.

task force to recommend changes to Neither of these licensing efforts, nor the tens of thousands of hours the NRC the NRC's inspection and oversight spent inspecting Fort Calhoun, led the agency to discover any of these many efforts, and then implement these safety problems.

changes as quickl,v as possible.

For two weeks irt June 2011,jloodirigon the Missouri River turned Nebraska's Fort Calhoun rtudear power plant into an island. The plant had already been shut downfo,* myriad safny prol>lems- mar1y datinir back lo its construction in the late 1960., and early 1970s.

Fort Calhoun's shutdown was not an isolated incident: The fact that there have been 52 year-plus outages demon-its two-and-a-half-year outage marked the fifty-second time a strates that U.S. reactors often operate while violating U.S. reactor remained shut down for longer than a year so the numerous safety requirements. These safety violations not owner could correct accumulated safety problems (see the only make reactors more vulnerable to accidents, but also table). In each of those cases, the reactor had been operating make them more likely to experience a Fukushima-scale w ith serious safety problems prior to the shutdown-problems disaster in the event of an accident.

that made a n accident more likely. Moreove1~ these 52 outages By closing the gap between what its safety regulations have cost ratepayers and shareholders billions of dollars. require and what U.S. plant owners actually do, the N RC The NRC's goal of preventing a Fukushima-scale accident would not only prevent another Fort Calhoun, it would also in this country must be accompanied by the goal of preventing strengthen its post-Fukushima reforms. And because year-another prolonged safety outage like that at Fort Calhoun. plus outages for safety fixes are costly, preventing another Year-Plus Nuclear Reactor Outages Date Date Outage Date Date Outage Outage Outage Length Outage Outage Length Reactor Began Ended (years) Reactor Began Ended (years)

Fermi Unit 1 10/5/66 7/18/70 3.8 Surry Unit 2 9/10/88 9/19/89 1.0 Palisades 8/11/73 10/1/74 1.1 Palo Verde Unit 1 3/5/89 7/5/90 1.3 Browns Ferry Unit 2 3/22/75 9/10/76 1.5 Calvert Cliffs Unit 2 3/17/89 5/4/91 2.1 Browns Ferry Unit 1 3/22/75 9/24/76 1.5 Calvert Cliffs Unit 1 5/5/89 10/4/90 1.4 Surry Unit 2 2/4/79 8/19/80 1.5 FitzPatrick 11/27/91 1/23/93 1.2 Three Mile Island Unit 1 2/17/79 10/9/85 6.6 Brunswick Unit 2 4/21/92 5/15/93 1.1 Turkey Point Unit 3 2/11/81 4/11/82 1.2 Brunswick Unit 1 4/21/92 2/11/94 1.8 San Onofre Unit 1 2/26/82 11/28/84 2.8 South Texas Project 2/3/93 5/22/94 1.3 Nine Mile Point Unit 1 3/20/82 7/5/83 1.3 Unit 2 Indian Point Unit 3 3/25/82 6/8/83 1.2 South Texas Project Unit 1 2/4/93 2/25/94 1.1 Oyster Creek 1.7 Indian Point Unit 3 2/27/93 7/2/95 2 .3 2/12/83 11/1/84 St. Lucie Unit 1 2/26/83 5/16/84 1.2 Sequoyah Unit 1 3/2/93 4/20/94 1.1 Browns Ferry Unit 3 9/7/83 11/ 28/84 1.2 Fermi Unit 2 12/25/93 1/18/95 1.1 Pilgrim 1.1 Maine Yankee 1/14/95 1/18/96 1.0 12/10/83 12/30/84 Peach Bottom Unit 2 4/28/84 7/13/85 1.2 Salem Unit 1 5/16/95 4/20/98 2.9 Fort St Vrain 6/13/84 4/11/86 1.8 Salem Unit 2 6/7/95 8/30/97 2.2 Browns Ferry Unit 2 9/15/84 5/24/91 6.7 Millstone Unit 2 2/20/96 5/11/99 3.2 Browns Ferry Unit 3 3/9/85 10.7 Millstone Unit 3 3/30/96 7/1/98 2.3 11/19/95 Browns Ferry Unit 1 3/19/85 6/12/07 22.2 Crystal River Unit 3 9/2/96 2/6/98 1.4 Davis-Besse 6/9/85 12/24/86 1.5 Clinton 9/5/96 5/27/99 2.7 Sequoyah Unit 2 8/22/85 5/13/88 2.7 LaSalle County Unit 2 9/20/96 4/11/99 2.6 Sequoyah Unit 1 8/22/85 11/10/88 3.2 LaSalle County Unit 1 9/22/96 8/13/98 1.9 Rancho Seco 12/26/85 4/11/88 2.3 D.C. Cook Unit 2 9/9/97 6/25/00 2.8 Pilgrim 4/11/86 6/15/89 3.2 D.C. Cook Unit 1 9/9/97 1 2/21/00 3.3 Peach Bottom Unit 2 3/31/87 5/22/89 2.1 Davis-Besse 2/16/02 3/16/04 2.1 Peach Bottom Unit 3 3/31/87 12/11/89 2.7 Fort Calhoun 4/9/11 12/21/13 2.7 Nine Mile Point Unit 1 12/19/87 8/12/90 2.6 SOURCE* UPDATED FROM LOCHBAUM 2006.

2 UNI.ON OF CONCERNED SCIENTISTS

safety problems reported by Fort Calhoun's owner during These year-plus outages

[ demonstrate that U.S.

the prolonged outage included:

Inade quate flood protection. NRC inspectors had

]

already determined in 2010 that measures designed to reactors often operate protect safety equipment in the auxiliary building and at while violating safety the intake structure from external flooding had not been adequately implemented as specified by the original requirements. safety studies. Workers identified additional deficiencies during the outage (Bannister 2011a). Furthermore, when the plant 's owner replaced the or iginal security system in Fort Calhoun would save ratepayers and sh areholders money.

1985, it left portions of the old system in place. Although Preventing financial meltdowns and avoiding reactor melt-the owner sealed the intake structure's walls up to the downs is a goal too good to pass up.

calculated flooding level to protect vital cooling water Just as it did for Fukushima, the NRC must formally pumps inside, it failed to seal areas where the old security examine the Fort Calhoun case, identify the lessons that system's cables penetrated the intake structure. As a should be learned, and make appropriate changes to its over-result, the safety-related water pumps could have bee n sight process to reduce the likelihood that safety problems damaged by flooding (Ba1111ister 2011b).

remain undetected- and uncorrected- for months or years.

Missing safety system parts. Fort Calhoun's owner in-stalled 32 seismically qualified General Electric electrical Safety Problems at Fort Calhoun relays in safety systems at the plant. Wor kers tested sev-en of these relays and three failed the tests. Workers then In a presentation to the NRC on March 27, 2013, Fort Calhoun's discovered the cause was a missing part. Further inquiries owner reported that 20,000 tasks had been completed between concluded that the relays were most likely missing this November 2012 and February 2013 and had approximately part when they were installed during the plant's original 5,000 other tasks to do before it could restart the reactor construction (Cortopassi 2013a).

(OPPD 2013). While many of these tasks involved preventive maintenance and routine inspections, some entailed Inadequate earthquake protection. Workers found correcting serious safety problems. that transmitters used to monitor reactor cooling water When a safety problem's severity rises above a fairly high pressure had been installed on an instrume nt rack that threshold, the plant owner must report it to the NRC. The was not designed to adequately protect them from In Ma.-ch 2013, Fort Calho,m*s owner reported that it had completed 20.000 tasks required b,v t/1(' NRC before the .-eacto.- could be restarted b11t still had approximately 5,000 1110,*e to do. Some ofthe tasks entailed correcting serious safety problems.

No More fukushimas; No More Fort Calhouns 3

movement during an earthquake. The owner informed this piping failed to comply with the piping code and the NRC that, "During a seismic event, the excessive therefore was not propedy qualified (Cortopassi 2012).

weight of these instrument racks could cause the racks Improperly grounded reactor protection system.

to fail," resulting in a reactor cooling water leak that Workers discovered that the voltage in the reactor could not be isolated, increasing the risk of nuclear protection system- which detects unsafe conditions core damage (Bannister 2012a). and initiates automatic safety system actions- was nearly Vulnerability to high-speed debris. In the event of 10 times higher than the design allowed. As a result, the a tornado, debris propelled by high winds can disable system might not initiate the automatic responses the essential safety equipment. Workers identified numerous pl:rnt's safety studies assumed would happen. Even potential sources of such debris, including removable worse, this unacceptable condition had been previously hatches on the intake structure, the exhaust stack for the identified and reported multiple times since 1993 but steam-driven auxiliary feedwater pump, the vent stack never corrected (Reinhart 2011).

and fill line for the emergency diesel generator's fuel oil tanks, the cable pull boxes for the raw water pumps, and

[

the exhaust stacks for the emergency diesel generators Workers discovered that (Cortopassi 2013b). some ofthe support beams Overloaded backup power source. Workers discovered that, in a situation where one of the two emergency diesel for the containment generators was unavailable, more equipment would be connected to the remaining emergency diesel generator structure were not than that generator could supply during certain types of accidents. The system designed to disconnect non-essential equipment from the emergency diesel generator during an accident would not perform properly during these types of accidents, and the overloaded generator properly designed to handle the weight they supported.

]

could fail to function (Bannister 2012b).

Safety pumps operated outside vendor limits. Work-Inadequately tested backup power source. In 1990, ers determined that, since 1996, the motors for the com-workers revised a test procedme for the emergency diesel ponent cooling water (CCW) pumps had been operating generators and no longer checked whether the plant's under conditions beyond those recommended by the fuel oil transfer pumps would automatically start and manufacturer. The CCW system supplies cooling water send fuel from the onsite storage tank to the generators. to reactor components that could contain radioactive This check, required by the react or's operating license, water (for example, reactor coolant pump lube oil and had not been performed for nearly a quarter of a century seal coolers, containment air cooling units, spent fuel (Bannister 2012c). pool heat exchanger). Motors operated outside the Overloaded support beam. Workers discovered that manufacturer's limits could fail during an accident some of the support beams for the containment structure (Bannister 2012e).

were not properly designed to handle the weight they Th is list summarizes only a handful of the safety prob-supported (Bannister 2012d). lems that eluded detection and correction at Fort Calhoun Inadequate piping qualifications. Workers discovered for years, subjecting the surrounding population to undue that chemical and volume control system (CYCS) piping elevated risk. The plant's problems covered a range of engi-had not been properly qualified for the stresses it could neering disciplines: electrical, mechanical, civil, and instru-experience during its lifetime. Among other factors, the ment and controls. They fell into several major safety areas, qualification was required to consider fatigue cycles- including fire protection, flood protection, and seismic that is, the number of times the water carried by the pip- design. In other words, the problems were programmatic ing goes from ambient temperature to reactor operating and pervasive, not isolated to a single plant department.

temperature and back again. These temperature changes The most recent of these problems dated to 1996, and cause the metal pipe walls to expand and shrink, which many dated back to when the plant was originally built. Thus, wears the piping out faster. Examination of two-inch- there were dozens, and sometimes hundreds, of opportunities diameter socket-welded fittings in the eves found that for workers and NRC inspectors to detect them before 2010.

4 UNI.ON OF CONCERNED SCIENTISTS

§

• ~

~

.,,§ i

E

]

z Senior execucives from che Fol'f Calhoun plane briefed NRC staffand commissioners several times (including here i11 Ju11e 2013) befol'e they were allowed to restar*t the reactor:

The NRC's Reactor Oversight Process inspectors' findings, then places the reactor into one of five Action Matrix columns. When the safety performance of a In May 1997 the Government Accountabilit y Office (GAO, reactor falls within the expected regime, the reactor is placed then cal1ed the General Accounting Office) issued a report in Column 1 and the NRC conducts only a baseline number titled Nuclear Regulation: Preventing Problem Plants Requires of inspections. As safety performance declines, the ROP man-More Effective NRC Action (GAO 1997). At the time, both dates supplemental NRC inspections. If safety performance reactors at New J ersey's Salem nuclear plant were mired in declines too much and a reactor falls into Column 5, the ROP year-plus outages and the NRC had identified 43 problems will trigger a shutdown until the owner fi xes the problems.

the owner had to correct before it could safely restart either The ROP Action Matrix for Fort Calhoun from the fourth unit. The GAO report stated that the NRC knew about 38 of quarter of 2000 (when the ROP program began) to the third the 43 problems before the Salem reactors were shut down, quarter of 2014 is shown in the figure on p. 6. The NRC moved and it knew about one of these problems for more than six years Fort Calhoun from Column 1 into Column 2 in the third prior to the shutdown. The GAO also documented that the NRC quarter of 2002, but later concluded that safety performance was aware of unresolved safety problems at the Millstone plant in Connecticut and the Cooper plant in Nebraska.

These findings prompted the GAO to conclude:

[

There were dozens, and

NRC has not taken aggressive enforcement action to force the licensees to fix their long-standing safety sometimes hundreds, of problems on a timely basis."

opportunities for workers "NRC allowed safety problems to persist because it was confident that redundant design features kept plants and NRC inspectors to inherently safe."

detect safety problems In response to criticism from the GAO and others, the at Fort Calhoun-NRC replaced its safety monitoring programs in April 2000 with its Reactor Oversight Process (ROP). The ROP evaluates a reactor's safety performance by combining 17 performance indicators (submitted quarterly by plant owners) with NRC opportunities that were missed.

]

No More fukushimas; No More Fort Calhouns 5

The NRC's ROP Action Matrix for Fort Calhoun, 2000-2014 2000 Q4 2001 Ql 2001 Q2 2001 Q3 ~

2001 Q4 2002 Ql 2002 Q2 2002 Q3 2002 Q4 2003 Ql 2003 Q2 ~

2003 Q3 1, 2003 Q4 2004 Ql 2004 Q2 b 2004 Q3 2004 Q4 2005 Ql 2005 Q2 2005 Q3 ~

2005 Q4 2006 Ql 2006 Q2 2006 Q3 ~

2006 Q4 1, 2007 Ql 2007 Q2 2007 Q3 II' 2007 Q4 2008 Ql 2008 Q2 ""

2008 Q3 2008 Q4 2009 Ql 2009 Q2 2009 Q3 2009 Q4 ~

2010 Ql 2010 Q2 2010 Q3 2010 Q4 ~

2011 Ql 2011 Q2 ~

2011 Q3 2011 Q4 2012 Ql 2012 Q2 2012 Q3 2012 Q4 2013 Ql 1, 2013 Q2 2013 Q3 2013 Q4 2014 Ql ~

2014 Q2 2014 Q3 0 2 3 4 5 ROP Column As a nuclear power plant's safety performance declines, the NRC moves it from Column I to Column 5 in t'1e Reactor Oversight Process Action Matrix. The NRC repeatedly moved Fort Cal'1oun back and forth in the matrix for over a decade until the agency decided the plant's problems were serious enough (Column S) to warrant a s'1utdown.

SOURCE* NRC N.0 6 UNI.ON OF CONCERNED SCIENTISTS

1

" NRC Commissioner William C. Ostendorff(leji) speaks with NRC Senior Resident I11speccor John Kirkland about repafrs needed at Fort Ca/'101111 while to11ri11g the plant during its .lO-month outage.

had improved and returned the reactor to Column 1. This or radiation release. At Fukushima, multiple problems caused happened again in the fourth quarter of 2003 and the third three reactoi-s to me lt down: the reactors lost off-site power, quarter of 2004. the backup generators located in the basements were damaged The NRC moved Fort Calho un into Column 3 in the when the basements flooded, floodwater disabled banks of second quarter of 2007 and the fourth quarter of 2007, but batteries that backed up the backup generators, and workers each time returned the plant to Column 2. When the NRC could not deploy portable pumps and generators in time.

again moved Fort Calhoun into Column 3 in the second The 1986 Chernobyl and 1979 T hree Mile Island accidents quarter of 2010, however, the plant subsequently slipped also occurred when numerous things went wrong.

into Column 4 and then into Column 5.

Thus, the ROP utterly failed to recognize the depth and Quite simply, the people breadth of the safety problems at Fort Calhoun until the third quarter of 2011. As noted above, all the safety problems sum-marized here existed at Fort Calhoun since at least 1996. They existed when the NRC returned Fort Calhoun from Column 2

[ ofNebraska faced unduly high risk for over a decade

]

to Column 1 on four occasions and when it returned Fort because the NRC did not Calhoun from Column 3 to Column 2 on two occasions.

These problems were so serious that Fort Calhoun could accurately evaluate safety not safely resume operation under NRC rules until each one was corrected, yet it had operated for over a decade with nil levels at Fort Calhoun.

of them. Quite simply, the people of Nebraska faced unduly high risk for over a decade because the NRC did not accu- Conversely, there have been cases where many things rately evaluate safety levels at Fort Calhoun. The ROP has went wrong and disaster was averted. For example, in 2002, clearly not fixed the problems identified by the GAO in 1997. workers at the Davis-Besse reactor in Ohio discovered that corrosion had caused a pineapple-sized hole in the reactor head, leaving only a thin steel cladding to contain the high-Preventing Another Fort Calhoun- pressure coolant. Once the reactor was shut down, workers and an American Fukushima discovered additional serious safety problems. Despite oper-ating with numerous safety problems, Davis-Besse avoided A key nuclear safety principle is "defense in depth." Reactors disaster because not all of its defense-in-depth barriers are designed so that no single problem will lead to a meltdown were compromised.

No More fukushimas; No More Fort Calhouns 7

Nevertheless, a reactor operating with pre-existing safety Bannister, D.J. 2012e. Licensee event report 2012-006, revision O,for problems is more vulnerable to disaster when another safety the Fort Calhoun Station. Omaha, NE: Omaha l'ublic Power District.

June 25. Online at http:// pbadupws.nrc.gov/ docs/ ML12l7/

problem arises. Fort Calhoun, before its reactor was shut down, ML12178A293.pdf was more likely to experience a Fukushima -scale accident Bannister, D.J. 2011a. L icensee event report 2011-003, revision 2,for because it was already operating with multiple pre-existing the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

safety problems. Pre-existing problems undermine defense in March l. Online at http://pbadupws.nrc.gov/ docs/ ML1206/

depth by reducing the number of things that must go wrong ML12061A224.pdf Bannister, D.J. 201 lb. Licensee event report 2011-003, revision 3,for to transform a near-miss into a nightmare.

the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

If the NRC's effort to prevent an American Fukushima is December 17. Online at http://pbadupws.nrc.gov/ docs/ ML1l3S/

to be successful, it must augment that with an effort to prevent ML11353055.5.pdf another Fort Calhoun. T he NRC responded to Fukush ima by Cortopassi, L.P. 2013a. Licensee event report 2013-008, revision O,for forming a task force t hat examined the accident and made more the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

June 7. Online at http://pbadupws.nrc.gov/ doc.s/ ML1315/

than 30 recommendations to better manage nuclear power ML13158A138.pdf plant risks. It is now in the process of implementing those Cortopassi, L.P. 2013b. Licensee event report 2013-009, revision O,for recommendations. the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

The NRC similarly needs to respond to Fort Calhoun by June 14. Online at www.nrc.gov/ site-help/ searcl1.cfm?q=ML13168A376.

forming a task force to determine how the agency and the plant Cortopassi, L.P. 2012. Licensee event report 2012-016, revision O,for the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

owner m issed- or dismissed- numerous lo,ngstanding safety September 17. Online at http://pbadupws.nrc.gov/ docs/ML1226/

p roblems for years despite thousands of ho urs of inspections. ML12262A31Zpdf The task force should recommend changes that will improve Government Accounting Office (GAO). 1997. Nuclear regulation:

the effectiveness and reliability of the NRC's inspection and Preventing problem plants requires more effective N1?C action.

oversight efforts. T he NRC then needs to implement these Washington, DC. May. Online at www.gao.gov/ products/

RCED-97-145.

changes as quickly as possible.

Lochbaum, D. 2006. Walking a nuclear tightrope: Unlearned lessons of year-plus reactor outages. Cambridge, MA: Un ion of Concerned REF ERENCES Scientists. September. Online at www.ucsusa.org/nuclear_power/

Bannister, D.J. 2012a. Licensee event report* 2012-010, revision O,for making-nuc/ear-power-safer/ who-is-responsible/walking-a-nuc/ear-the Fort Calhoun Station. Omaha, NE: Omaha Public Power District. tightrope.htm l#. VOYn5c8o670.

August 3. Online at http://pbadupws.nrc.gov/ docs/ ML1221/

Nuclear Regulatory Commission (NRC). No date. ROP historical perfor-ML12219AOJO.pdf mance from previous quarters. Rockville, MD. Onl ine at www.nrc.

Bannister, D..J. 2012b. Licensee event report 2012-011, revision O,for gov/ NRR/0 VE RSIGI-IT/ ASSESS/ prevqtr.html.

the Fort Calhoun Station. Omaha, NE: Omaha Public Power District. Omaha Public Power District (OPPD). 2013. Fort Cal/10un Station driving Augus t 6. Online at http://pbadupws.nrc.gov/ docs/ ML1222/ through restart. Omaha, NE: Omaha Public Power District. March 27.

MLI 2220A 16Zpdf Online at https://adamswebsearch2.nrc.gov/ webSearch2/ main.jsp?

Bannister, D.J. 2012c. Licensee event report 2012-005, revision O,for AccessionNumber=ML13093A473.

the F'ort Call1oun Station. Omaha, NE: Omaha Public Power District.

Reinhart, J.A. 2011. Licensee event report 2011-002, revision 1,for the April 23. Online at http:// pbadupws.nrc.gov/ docs/ MLl22S/

Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

ML122SOA189.pdf July 27. Online at http://pbadupws.nrc.gov/ docs/ MLJJ20/

Bannister, D.J. 2012d. Licensee event report 2012-014, revision O,for MLJ/2081990.pdf the Fort Calhoun Station. Omaha, NE: Omaha Public Power District.

September 10. Online at http://pbadupws.nrc.gov/ docs/ ML122S/

ML12255A038.pdf

[

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NATIONAL HEADQUARTERS WASHINGTON, DC, OFFICE WEST COAST OFFICE MIDWEST OFFICE

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Unionof d * *

[ Concerne Scientists EXECUTIVE

SUMMARY

The NRC and Nuclear Power Plant Safety in2014 Tarnished Gold Standard OUR FIFTH ANNUAL REPORT CARD The Nuclear Regulatory Commission (NRC) often claims to r epresent the gold The NRC often claims to be the gold standard for nuclear powe r plant safety regulation and oversight (Macfarlane standard fm* nt1cl<'ar power plant safety 2013; Magwood 2013). Ample evidence, including the summaries of positive outcomes achieved by the NRC in this series of annual reports, suggests much regulation and ovcrsiF(ht. Ample evidence validity to these claims. One cannot count the number of nuc lear disasters averted suggests much validity to these claims.

by the NRC's effective regulatory performance, but one can generally count on One ca1111ot count the number of nuclear the NRC to be an effective regulator. The NRC has done much to earn the gold disasters averted b_y the NRC's effective standard label.

reg11/atory pcrfonnance, but one Chapter 4 of this report describes how the NRC conducted two extensive can generally count on the NRC reassessments of its reactor oversight process- not in respon se to an accident demonstrating its inadequacy or to criticism suggesting an inadeq1.1acy, b1.1t as to be an effective regulator.

a proactive measure aimed at enhancing the effectiveness and efficiency of the existing process. Chapter 4 also descr ibes how a decade ago t he NRC recognized Hur thr NRC'sgo/d standard is it had an aging work force and developed formal programs to retain as m1.1ch tarnished. For the past decade, they have tribal knowledge as possible before its retirees hit the golf courses and beaches been improperly withholding dornm ents in their golden years. Such proactive actions enable the NRC to retain the gold about s,1fet_y problems, have subj('ctcd standard label.

engineers who voicrd safr(Y concrms to Chapters 2 and 3 of this report describe how the numbe1* and severity of near misses at nuclear power plants have been steadily declining since 2010 repeated investigations ofalleged (Table 1, p. 2), again consistent with the NRC being an e ffective reg1.1lator.

but unsubstantiated wrongdoing, and l1a\ll' been usi11gncm11nifbrm answer kl.'ys to grade standardized tt*sts administered via ics reactor oversight process.

If the NRC truly is the gold standard, it must restore the luster and prevenl the tarnish .from recurring z

~

The Mills1011e Power Station i11 Waterfiird. CT, which experi<'nced cwo .1elf-i11flic1ed 11ear misses in 2014 whc11 recent maintenance and modifications introduced proble ms that reduced .'iafety margins.

TABLE 1. Near Misses 2010 to 2014 Total Number Near Near Near Near Near of Near Misses in Misses in Misses in Misses in Misses in Reactor Misses 2010 2011 2012 2013 2014 1 Arkansas Nuclear One Unit 1 2 1 1 2 Arkansas Nuclear One Unit 2 2 1 1 3

-Braidwood Unit 1 2 1 1 4

5

-Braidwood Unit 2 Browns Ferry Unit 1 2

1 1 1 1

6 Browns Ferry Unit 2 1 1 7 Browns Ferry Unit 3 1 1 8 Brunswick Unit 1 1 1 9 Brunswick Unit 2 2 1 1 10 Byron Unit 1 1 1 11 Byron Unit 2 2 1 1 12 Callaway 1 1 13 Calvert Cliffs Unit 1 2 1 1 14 Calvert Cliffs Unit 2 2 1 1 15 Catawba Unit 1 3 1 1 1 16 Catawba Unit 2 1 1 17 Clinton 1 1 18 Columbia 3 3 19 Cooper 1 1 20 Crystal River Unit 3 1 1 21 Davis-Besse 1 1 22 23 24

-Diablo Canyon Unit 2 Farley Unit 1 Farley Unit 2 2

1 1

1 l

l l

25 Fermi Unit 2 1 1 26 Fort Calhoun 4 1 2 1 27 Grand Gulf 1 1 28 H.B. Robinson 2 2 29 Joseph M. Farley Unit 2 1 1

.30 LaSalle Unit 1 1 1 31 32

- LaSalle Unit 2 Millstone Unit 2 2 1

1 1

1 33 Millstone Unit 3 2 2 2 UNI.ON OF CON CERN ED SCIENTISTS

TABLE 1. Near Misses 2010 to 2014 (continued)

Total Number Near Near Near Near Near of Near Misses in Misses in Misses in Misses in Misses in Reactor Misses 2010 2011 2012 2013 2014 34 North Anna Unit 1 1 1 35 North Anna Unit 2 1 1 36 Oconee Unit 1 1 1 ,_

37 Oconee Unit 2 1 1 38 Oconee Unit 3 1 1 39 Oyster Creek 1 1 40 Palisades 3 2 1 41 Palo Verde Unit 1 1 1 42 Palo Verde Unit 2

-- 1 1 43 Palo Verde Unit 3

- 1 1 44 Perry 2 1 1 45 Pilgrim 2 2 46 River Bend 2 1 1 47 San Onofre Unit 2 1 1 48 San Onofre Unit 3 1 1 49 Shearon Harris 2 1 1 50 Surry Unit 1

- - 1 1 51 Susquehanna Unit 2 1 1 52 Turkey Point Unit 3 1 1 53 Wolf Creek 4 1 1 2

'im ID:) ml m> ml w m The overall number ofnear misses continues to decline each year, as does the number ofaffected sites and the severity ofevents.

SOURCE: ucs_

But Chapter 5 reveals the gold standard to be tarnished. how the NRC has been using nonuniform answer keys to For the past decade, the NRC has been improperly withholding grade standardized tests administered via its reactor over-documents, including many about safety problems. By doing sight process (Table 2, p. 4), yielding numerica] outcomes less so, the NRC deprived the public oflegal rights for regulatory predictable trhan fluctuating gold prices. By improperly with-decision-making and painted a misleading picture of nuclear holding many safety problem reports and jiggling the grading safety. Chapter 5 also describes how two NRC engineers who of other safety problems, the improving trends may be more did their duties and voiced safety concerns were subjected fabrication than fact. If the N RC truly is the gold standard to repeated investigations of alleged but unsubstantiated of nuclear re gulators, it must restore the luster by removing wrongdoing, sending a very clear message throughout the this tarnish and preventing it from recurring.

agency that "silence is golden." Finally, chapter 5 explains The NRC and Nudea r Power Plant Safety in 201 4 3

TABLE 2. Seven Cornerstones of the Reactor Oversight Process Initiating Condit ions that. if not properly controlled, requ ire the plant's emergency equipment to maintain safety.

Events Problems in this cornerstone include improper control over combustible materials or welding activities, causing an elevated risk of fire: degradation of piping . raising the risk that It will rupture: and improper sizing of fuses. raising the ri sk that the plant w ill lose electrical power.

Mitigating Emergency equipment designed to limit the impact of init iating events. Problems in this cornerstone include Systems ineffective maintenance of an emergency d iesel generator, degrading the ability to provide emerg ency power to respond to a loss of offsite power: inadequate repair of a problem with a pump in the emergency reactor-core cooling system, reducing the reliability of cooling during an accident: and non-conservative calibration of an automatic temperature set point for an emergency ventilation system. delaying its startup longer than safety studies assume.

Barrier Integrity Multiple forms of containment preventing the release of rad ioactive material into the environment. Problems in this cornerstone include foreign material in the reactor vessel. which can damage fuel assemblies: corrosion of t he reactor vessel head; and malfunction of valves in piping t hat passes t hrough conta inment walls.

Emergency Measures intended to protect the public if a reactor releases significant amounts of rad ioactive material.

Preparedness Problems in this cornerstone include emergency sirens within 10 miles o f the plant that fail to work:

and underestimation of the severity of plant conditi ons during a simulated or actual accident. delaying protective measures.

Public Radiation Design features and administrative controls that limit public exposure to radiation. Problems in this Safety cornerstone include improper calibration of a radiation detector that monitors a pathway for the release of potentially contaminated air or water to the environment.

Occupational Design features and administrative controls that limit the exposure o f plant workers to rad iation. Problems Radiation Safety in t his cornerstone include failure to survey an area properly for sources of radiation, causing workers to receive unplanned exposures; and incomplete accounting of individuals' radiation exposure.

Security Protection against sabotage that aims to release radioactive material into the environment. which can include gates. guards, and guns. After 9/11, the NRC reduced the discussion of this cornerstone in the public arena.

The NRC's Reaction Oversight Process f eatures seven cornerstones ofreactor safety to help inspectors detect problems before they become more serious.

SOURCE. WWW NRC.GOV/REACTORS/OPERA TING/OVERSIGHT/ROP*DESCR/PTION.HTML

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• FIND THIS DUCUMENT ONLINE: www.ucsusa.org/N RC2014 The Union ofConcerned Scientists pt1ts rigorous, independent science to work to solve our planet's most pressing problems. Joining with citizens across the country, we combine technical analysis and effective advocacy to create innovative, practical solutions for a healthy, safe, and sustainable future.

NATIONAL HEADQUARTERS WASHINGTON, DC, OFFICE WEST COAST OFFICE MIDWEST OFFICE

'l\vo Hrattle Square 1825 I< St. NIN, Suite 800 500 12th St., Suite 340 One N. LaSalle St., Suite 1\104 Cambridge, MA 02138-3780 Washington, DC 20006-1232 Oakland, CA 94607-4087 Chicago, IL 60602-4064 Phone: (617) 547-5552 Phone: (202) 223-6133 Phone: (510) 843-1872 Phone: (312) 578-1750 Fax: (617) 864-9405 Fax: (202) 223-6162 Fax: (510) 843-3785 Fax: (312) 578-1751 w1m: www.ucsusa.org 0 PRIN'FED ON RtCYCLEO PJ\Pt:R USJNG VEGETAHl,t-.*8ASED INK~ ((:' MARCH 2015 UNION OF COi\Ct-~RNJ::.D 1i,CJEX'l ISTS