ML15110A476

Foe Reply in Support of Motion to Supplement - Ex.1

ML15110A476
Person / Time
Site:	Diablo Canyon
Issue date:	04/17/2015
From:	Ayres Law Group, Friends of the Earth, Pacific Gas & Electric Co
To:	NRC/OGC, US Federal Judiciary, US Court of Appeals for the District of Columbia Circuit
Creedon, Meghan
Shared Package
ML15110A474	List: ML15110A475 ML15110A476 ML15110A477
References
Download: ML15110A476 (36)

Text

USCA Case #14-1213 Document #1547998 Filed: 04/17/2015 Page 1 of 36 EXHIBIT 1 PETITIONERS REPLY TO RESPONDENTS AND INTERVENORS RESPONSES TO PETITIONERS MOTION TO SUPPLEMENT THE CERTIFIED INDEX OF THE RECORD No. 14-1213 (D.C. Cir.)

Differing Professional Opinion - Appeal of Dr. Michael Peck From Case File for DPO-2013-002 (Sep. 9, 2014) (ML14252A743)

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Continued Item 11 NRC Form 690 Page 1

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Continued Item 11

2. The Panel Report did not provide sufficient detail to support the conclusion that the licensees actions were consistent with agency statutory requirements. The DPO detailed specific examples of the agencys failure to enforce certain regulatory and statutory requirements. The Panel Report responded to these detailed examples with general statements that regulatory requirements and safety objectives were satisfied.

Background

The DCPP seismic design and local geology is complex. However, the facility design control (10 CFR 50, Appendix B), License fidelity (10 CFR 50.59 and 10 CFR 50.71(e)), and operability (DCPP Technical Specification) issues raised in the DPO were not overly complex. These processes are well understood and routinely verified as part of the NRC Light Water Reactor Inspection Program and the Reactor Oversight Process.

In November 2008, PG&E reported discovery of a new line of epicenters located about a mile offshore from the DCPP.1 The licensee stated that if this line of epicenters represented an earthquake fault, then the resulting ground motion would be bounded by the DCPP seismic design bases established by the Long Term Seismic Program (LTSP). The licensee committed to characterize the potential fault and evaluate the effect on plant structures, systems, and components (SSCs). This line of epicenters became known as the Shoreline fault.

In April, 2009, the NRC Office of Nuclear Reactor Regulation (NRR) released a preliminary review of the Shoreline fault. 2 This analysis concluded that ground motion that may be produced by the Shoreline fault would be within the plant seismic design bases (LTSP). NRC personnel, myself included, presented the results of this preliminary review at multiple public meetings held during the subsequent two years.

In September 2010, the NRC and PG&E held a public seismic workshop in San Luis Obispo, California. During the workshop, a PG&E seismologist presented the results of deterministic and seismic hazard characterization of the Shoreline fault. At the end of the presentation, I asked how the new ground motions compared to the facility SSE. The PG&E seismologist did not answer my question. The seismologist stated that LTSP established the facility seismic design basis. After the workshop, I reviewed the facility SSE as presented in the FSARU. I found that the seismic design basis documented in the FSARU was considerably different than both PG&E and the NRC personnel had described at the pervious public meetings. The FSARU stated that the LTSP was explicitly not part of the seismic design basis. I also found that the Shoreline fault deterministic ground motions, as presented at the workshop, were about 70 percent greater than those described in the facility SSE safety analysis.

Per Inspection Procedure IP 71111.15,3 an operability evaluation was required because the new information called into question if the seismic design basis, as established by General Design 1

NRC Event 44675, Offsite Notification and Media Briefing due to Potential Discovery of Off Shore Fault near Plant, November 21, 2008.

2 Diablo Canyon Power Plant, Unit Nos. 1 and 2 - NRC Preliminary Review of Potential Shoreline Fault, April 8, 2009 (ML090930459).

3 Inspection Procedure 71111.15, Operability Determinations and Functionality Assessments (ML112010663), If operability is not justified then determine impact on any TS limiting condition for operation (LCO).

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Continued Item 11 Criteria (GDC) 2, was still satisfied.4 To be considered operable, technical specification required SSCs must be capable of performing the required safety functions, as described in the safety analyses, at the higher seismic loadings. PG&E maintained that operability evaluation was not required because the new ground motions were within the bounds of the LTSP.

In November 2010, I presented my findings to Region IV management and the NRR project manager (PM). At this meeting the deputy director of Division of Reactor Projects (DRP) took an action to request PG&E to formally evaluate the operability of plant SSCs. PG&E again refused, stating that the LTSP established the seismic design basis for the facility.

I concluded that PG&E would likely not be successful demonstrating operability based on my previous experience with DCPP reactor head replacement inspections. These inspections identified that some reactor coolant pressure boundary and reactor head structural components failed to meet the American Society of Mechanical Engineers (ASME) Code5 acceptance limits when evaluated against the existing double design earthquake (DDE) or SSE loads.6 PG&E subsequently obtained an amendment to the Operation License allowing use of higher seismic damping values in the Code calculations.7 This inspection revealed that insufficient Code margin was available to accommodate the higher loading represented by the Shoreline fault.

In December 2010, I reported back to the DRP deputy director that PG&E had not preformed the requested operability evaluation. The deputy director encouraged me to drop the issue. The deputy director suggested that, as an option, I could prepare a white paper detailing the concern.

In January 2011, PG&E submitted the completed reevaluation of the local geology on the DCPP Docket.8 This report included deterministic evaluations concluding that three local faults, the Shoreline, Los Osos and San Luis Bay, were each capable of generating significantly greater ground motion than was used to establish the facility DDE/SSE.

In February 2011, I submitted a white paper to Region IV management.9 The white paper described the facility seismic design bases and the extent the new ground motions exceeded the limiting values used the DDE/SSE safety analysis to seismically qualify plant SSCs. I included recommendations to initiate enforcement action against PG&E. These recommendations included 4

NRC Inspection Manual, Part 9900: Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded Or Nonconforming Conditions Adverse to Quality or Safety (ML073531346), Section C.1, Relationship Between the General Design Criteria and the Technical Specifications, stated that the failure to meet a General Design Criteria in the CLB should be treated as a degraded or nonconforming condition and, therefore, the technical guidance in this document is applicable. The Diablo Canyon CLB established the DDE as the GDC 2 SSE.

The new ground motions exceeded the SSE ground motions described in the FSARU 5

American Society of Mechanical Engineers Boiler and Pressure Vessel, Code,Section III, required per 10 CFR 50.55a.

Meeting Code acceptance limits ensures the integrity of the reactor coolant pressure boundary following earthquakes and accidents 6

Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2009005 And 05000323/2009005, February 3, 2010 ( M100341199) 7 Diablo Canyon Power Plant, Unit Nos. 1 And 2 -Issuance Of Amendments Re: Revision To Final Safety Analysis Report Update Section 3.7.1.3, "Critical Damping Values" (TAC NOS. ME4056 AND ME4057) (ML102530443) 8 Report on the Analysis of the Shoreline Fault Zone, Central Coast California to the NRC, January 7, 2011 (ML110140400) 9 White Paper, Resident Inspectors Recommendation for Regulatory Disposition of the Failure of Pacific Gas & Electric to Perform an Operability Evaluation Following Discovery of the Shoreline Fault, February 2, 2011, attached to e-mail to Geoff Miller,

Subject:

ACT: Diablo Canyon - Recommendation for Regulatory Disposition.

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Continued Item 11 a potential greater than green finding associated with the failure of PG&E to evaluate and disposition SSC operability (10 CFR 50, Appendix B) and an escalated traditional enforcement violation (10 CFR 50.9) after PG&E provided incomplete and inaccurate information concerning the facility seismic design bases. This incomplete and inaccurate information was used by the NRR PM for the agencys conclusions presented in the April 2009 letter.

In March 2011, a meeting was held at Region IV to discuss the white paper recommendations.

The branch chief from the NRR Division of Operating Reactor Licensing, the NRR PM and DRP management attended the meeting. A consensus was reached that PG&E had not evaluated the new seismic information against the facility design bases. The DRP division director expressed concern that enforcement action would conflict with the NRC position communicated in the April 2009 NRR letter.10 To address this concern, I drafted a concurrence Task Interface Agreement (TIA) letter documenting agreement between NRR and Region IV that PG&E was required to evaluate the new seismic information against the facility design bases, including the DDE/SSE.11 The failure of the licensee to perform an operability evaluation was documented as an unresolved item (URI) in the DCPP inspection report.12 Between December 2010 and June 2011, the NRC and PG&E held several public meetings to discuss how the new seismic information would be incorporated into the DCPP License. PG&E proposed using the Hosgri Evaluation (HE) methodology for the facility SSE. The HE described the plant response to a postulated 7.5 Magnitude earthquake on the Hosgri fault. The HE used different assumptions, methodology and acceptance limits than the existing DDE/SSE. The CLB described the HE as a response to a NRC question raised during original plant licensing. The HE bound the higher ground motions identified in the PG&E reevaluation of the local geology.

In October 2011, PG&E submitted License Amendment Request (LAR) 11-05 to designate the HE as the DCPP SSE.13 Also, in October 2011, PG&E concluded that all plant SSCs were operable in response to the URI and TIA.14 However, the licensee only evaluated the new ground motions against the HE. The licensee stated that NRC operability policy provided for use of the HE as an alternative method.

Based on using the HE alternative method, PG&E argued that the new ground motions did not have to be directly evaluated against the DDE/SSE safety analysis or acceptance limits. Based on 10 Diablo Canyon Power Plant, Unit Nos. 1 and 2 - NRC Preliminary Review of Potential Shoreline Fault, April 8, 2009 (ML090930459). Letter stated that the LTSP established the seismic design bases 11 Task Interface Agreement - Concurrence on Diablo Canyon Seismic Qualification Current Licensing and Design Basis, August 1, 2011 (ML112130665) 12 Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011002 and 05000323/2011002, May 11, 2011, Unresolved Item: 05000275; 323/2011002-03, Requirement to Perform an Operability Evaluation Following Receipt of New Seismic Information. URI updated in Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011003 And 05000323/2011003, August 10, 2011, Discussed URI 05000275;05000323/2011002-08, Requirement To Perform An Operability Evaluation Following Receipt of New Seismic Information (Section 4OA2).

13 License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake" October 20, 2011 (ML11312A166).

14 PG&E Notification: 50086062, Type: DA Work Type: EVAL AANS,

Description:

LTCA-Ident. of Seis. Lineament Offshore.

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Continued Item 11 the PG&E operability evaluation, the NRC closed the URI and issued a violation associated with the failure to evaluate operability after initially developing the new seismic information.15 I disagreed that the HE satisfied NRC criteria for use as an alternative method for operability. I included a violation with DCPP Inspection Report 2011-05 to address PG&Es inadequate operability evaluation. Region IV management did not accept my recommended violation. The licensee stated that comparison of the new seismic information directly against the DDE/SSE safety analysis would result in exceedances. In other words, operability could not be demonstrated by comparing the new seismic information with the GDC 2 design basis and safety analysis. This was a concern because the HE, while bounding for ground motion, was not bounding for the seismic qualification of technical specification required SSCs.16 I documented my concerns using the NRC non-concurrence process.17 I included a detailed technical discussion addressing why the PG&E operability evaluation failed to meet the NRC standard. I expected Region IV to agree with the technical argument and issue the recommended violation. I also expected PG&E to follow up with a request for regulatory dispensation in the form of a waiver (10 CFR 50.12) and Code relief (10 CFR 50.55a) due to the lack of margin in the existing DDE/SSE safety analysis. The alterative required PG&E to perform a plant technical specification shut down pending disposition of the non-conforming safety analysis.

In response to the technical discussion in the non-concurrence, the agency stated:

the seismic CLB did not provide a way to evaluate new information that becomes available. Therefore, the licensee has proposed a methodology to perform the full operability evaluation to the NRC as a license amendment request, and the staff is evaluating the best way to proceed.

the complete operability evaluation cannot be made by the licensee without the NRC agreeing on the correct way to perform the evaluation, what calculation method and design values are appropriate for the new data, and what plant capability must be demonstrated by this evaluation.

The NRC will not ask the licensee to use the new ground motion input data in the Design or the Double Design Earthquake (SSE) evaluations because the new ground motion data does not match the assumptions in those analyses. Attempting to do so would create a numerical result that is not technically justified.

The staff concluded the revised operability determination provided an initial basis for concluding a reasonable assurance that plant equipment would withstand the potential effect of the new vibratory ground motion.

Rather than addressing the specific technical issues presented in the non-concurrence, Region IV presented an argument that PG&E did not have to meet technical specification operability requirements. Region IVs apparent argument was that operability cannot be demonstrated against the current safety analysis; therefore operability may be deferred until the NRC approves a method (LAR 11-05) that would have a successful result.

This was a concern because NRC policy did not provide for continued reactor operation outside of the bounds of limiting safety analysis unless the licensee clearly demonstrated SSC operability.

15 NCV 05000275; 05000323/-2011005-02, Failure to Perform an Operability Determination for New Seismic Information (Section 1R15.2), Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011005 and 05000323/2011005 , February 14, 2012 (ML12040843).

16 Detailed examples were provide in DPO 2013-002 17 Non-Concurrence NCP-2012-001, DCPP IR 2011-05 (ML12045843)

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Continued Item 11 NRC policy did not provide for an initial basis for operability or deferment until the License is amended. Continued reactor operation was only permitted after SSCs were demonstrated operable at that point in time. Plant SSC are considered inoperable, and the associated technical specification Limiting Condition for Operation not met when a nonconforming or unanalyzed condition results in an SSC unable to perform its specified safety function as described in the safety analysis.18 In February 2012, the NRC concluded that LAR 11-05 (requested to adopt the HE for the facility SSE) would not be accepted for review.19 The staff rejected the LAR because:

1) The methodologies and acceptance limits for SSCs using HE differ from that specified in Standard Review Plan (NRC acceptance criteria for a facility SSE).

2) PG&E had not completed a reevaluation of the reactor coolant system for the seismic and LOCA loads (the HE didnt meet ASME Code requirements for the SSE).

3) PG&E did not provide a peer reviewed seismic probabilistic risk assessment.

4) Concerns about use of a seismic margins assessment for operability evaluations.

In October 2012, PG&E withdrew LAR 11-05 at the NRCs request.20 Also, in October, the NRR PM provided PG&E written direction to update the FSARU to include the Shoreline scenario as a lesser included case under the HE.21 The PMs action essentially established the LTSP and HE as the de-facto SSE, circumventing the license amendment process per 10 CFR 50.90,22 and bypassing the required public notice and hearing opportunities required for a change to the Operating License per 10 CFR 50.91.23 The PM justified this action by stating:

As documented in RIL 12-01, the NRC staff's assessment is that deterministic seismic-loading levels predicted for all the Shoreline fault earthquake scenarios developed and analyzed by the NRC are at, or below, those levels for the Hosgri earthquake (HE) ground motion and the long term seismic program (LTSP) ground motion.

The HE ground motion and the LTSP ground motion are those for which the plant was evaluated previously and demonstrated to have reasonable assurance of safety. Therefore, the staff has concluded that the Shoreline scenario should be considered as a lesser included case under the Hosgri evaluation and the licensee should update the final safety analysis report (FSAR), as necessary, to include the Shoreline scenario in accordance with the requirements of 10 CFR 50.71(e).

18 NRC Inspection Manual, Part 9900: Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety (ML073531346), Sections 3.8, 3.10

& 6.1 19 FOIA/PA NO: 2014-0065 (Group B) (ML13354B992) 20 Diablo Canyon Power Plant Units 1 and 2 - Withdrawal of an Amendment Request, October 31, 2012 (ML12289A076) 21 Diablo Canyon Power Plant Units 1 and 2 - NRC Review of Shoreline Fault(ML120730106) 22 NRR Office Instruction LIC-100, Revision 1, Control of Licensing Bases for Operating Reactors, Section 2.1.5.5 10 CFR 50.90, License Amendments (ML033530249) 23 See the Perry Decision, Commission Memorandum and Order CLI 96-12 NRC Form 690 Page 6

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Continued Item 11 As discussed in detail in the DPO, demonstration to have reasonable assurance of safety was not among the criteria used by NRC to determining if an amendment to the Operating License was required.24 In July 2013, I submitted DPO-2013-002, Differing Professional Opinion Involving Seismic Issues at DCPP. This DPO identified three concerns:

1) Incorporating the Shoreline scenario into the FSARU required prior NRC approval in the form of an amendment to the Operating License.

2) Region IV failed to enforce DCPP Technical Specification requirements for a plant shutdown after the licensee inadequately operability evaluation.

3) The Agency failed to adequately disposition the updated seismic information associated with San Luis Bay and Los Osos earthquake faults.

In May 2014, the DPO Panel Report was issued. I agreed with the Panels conclusion that issues raised in the DPO did not result in a significant or immediate safety concern. I also agreed that the potential ground motions from the nearby faults would not exceed the levels of ground motion considered during the licensing of the plant. However, I disagreed with the Panels other conclusion:

1) An amendment to the Operating License was not required for the new seismic information.

2) A lack of formal regulatory guidance exists for evaluating new information on natural hazards.

3) The licensee adequately demonstrated SSC technical specification operability.

Original Diablo Canyon Power Plant Seismic Design and Licensing Bases An understanding of the facility licensing bases is needed before a effective review of the DPO Panel conclusion can be performed.

The FSAR (as amended) served as the principal reference document to support the PG&E Part 50 DCPP license application. The FSAR described the methods PG&E used to confirm that applicable NRC regulations were met and contained the technical information required by 10 CFR 50.34.

This technical information included safety analyses that presented the design bases and the limits on operation for plant SSCs. 10 CFR 50.34(b) specifically required the FSAR to include safety analyses that demonstrated that the principal design criteria for the facility (GDCs) were met. This included the design basis and the relationship of the design bases to these principal design criteria (GDCs).

10 CFR 50.2 defined design bases as that information which identifies the specific functions to be performed by a facility SSC and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. 10 CFR 50.2 design bases included the bounding conditions under which SSCs must perform design bases functions, including protection against 24 NRC criteria used to determining if an amendment to the Operating License is required is found in 10 CFR 50.59.

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Continued Item 11 natural phenomena. For seismic, the design bases functional requirements were derived primarily from the principal design criteria contained in GDC 2 (the minimum standards set by Part 50, Appendix A) and NRC regulations that imposed functional requirements or limits on the plant design (10 CFR 100, Appendix A). These 10 CFR 50.2 design bases were a subset of the original licensing bases.

The original DCPP FSAR, including the 10 CFR 50.2 design bases, were presented in accordance with 10 CFR 50.34(b)25 and were reviewed by the NRC in connection with granting the original license. These safety analyses (license application, FSAR Amendment 85) became the original plant licensing bases when the NRC approved the facility Operating License.

Ive included exerts of the FSAR (license application, Amendment 85) in Appendix A. The original seismic licensing bases may be summarized as:

The seismic design basis functional requirements were established by GDC 226 and 10 CFR 100, Appendix A. The DDE safety analysis (FSARU Sections 2.5, 3.7, 3.8, 3.9, 3.10, and 5.2) demonstrated that the GDC 2 and Part 100, Appendix A, SSE design bases functional requirements were satisfied.

The earthquake design bases were defined as the DE and DDE (equivalent to the Part 100, Appendix A, operational basis earthquake and SSE).

The GDC 2 safety analysis (FSAR 2.5.2.9) determined that the DDE was the maximum earthquake potential for the facility (considering all faults within 75 miles of the site). This safety analysis was consistent with the requirements 10 CFR 100, Appendix A. The Hosgri was not considered a capable27 fault and excluded from the GDC 2 safety analysis.

The HE was prepared to answer a NRC question. The HE was not included in the 10 CFR 50.34 safety analyses (FSAR Section 2.5) because the HE did not implement a regulatory requirement per 10 CFR 50.34. PG&E maintained the HE, a beyond design bases event, as a licensing bases commitment.28 PG&E only committed to seismically qualify plant SSCs (needed to function for the SSE per Safety Guide 29, Seismic Design Classification) for the DDE.29 Some plant SSCs were also qualified for the HE. In many cases the seismic qualification of plant SSCs were more limited 25 Also consistent with PG&Es commitment to Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition) 26 FSAR stated that PG&E met GDC 2 (1997). However, Letter, from A. Giambusso, Director of Licensing, Atomic Energy Commission (AEC), to F.T. Searls, Pacific Gas and Electric, dated August 13, 1973, committed PG&E to address any deviations or exceptions taken to GDC 2 (Part 50, Appendix A, 1971). Letter: F. J. Miraglia, Division of Licensing, US NRC, from P. A. Crane, Pacific Gas and Electric, CHRON 131464, Description of PG&Es compliance with the requirements 10 CFR 20, 50, and 100, dated September 10, 1981, included that DCPP seismic design bases did not include any exceptions to GDC 2 (Part 50, Appendix A, 1971).

27 Capable defined per 10 CFR 100, Appendix A. At the time of OL, NRC and PG&E disagreed on the capability of the Hosgri fault (see DCPP SSER 7).

28 Regulatory Guide 1.186, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, endorses use of NEI 97.04, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases Appendix B, for providing examples and guidance acceptable to the staff for providing a clearer understanding of what constitutes design bases information. NEI 97.04, Appendix B stated that design bases are explicitly tied to regulatory requirements, primarily the GDCs, and implemented by the 50.34 safety analyses. The HE does not implement a regulatory requirement or GDC and this not included within the GDC 2 design bases.

29 Set of SSCs listed in Safety Guide 29 (Regulatory Guide 1.29, Seismic Design Classification), required to remain functional following a SSE.

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Continued Item 11 for the SSE/DDE than the HE. As described in the DPO, this was based on differences in the assumptions, methods, and acceptance criteria used in the two analyses.

Diablo Canyon Power Plant Current Licensing Basis FSARU, Revision 20, was the current FSARU when the DPO was written. The CLB seismic and design bases were very similar to the original licensing bases. In summary, the CLB:

The DDE and supporting safety analysis satisfied the requirements of GDC 2 and were equivalent to the SSE described in 10 CFR 100, Appendix A.

The licensee committed to ensure the plant SSCs listed in Regulatory Guide 1.29 (Seismic Design Classification) will remain functional following the DDE/SSE.

The HE was an answer to an NRC question during original plant licensing. Regulatory Guide 1.29 does not apply to the HE.

FSARU Section 3.7.6 established the HE shutdown path. Unlike the DDE/SSE (GDC 2), the HE did not assume a coincidental accident or fire. This section described the SSCs qualified for the Hosgri earthquake.

As required by 10 CFR 50.55a, PG&E demonstrated that the combined accident and DDE/SSE loads did not exceed ASME Code acceptance limits for the reactor coolant pressure boundary.

PG&E performed ASME Code calculations for the HE. However, PG&E did not include accident loads in these calculations. HE Code calculations were not required by NRC regulations. PG&E performed these calculations as part of a licensing bases commitment.

The HE was not tied to meeting a regulatory requirement (GDC, Part 100, etc.). Because HE was not part of the design bases, the licensee was not required to include a 10 CFR 50.34 safety evaluation in the FSARU.30 LTSP was explicitly excluded from the seismic design bases. PG&E maintained a licensing bases commitment to evaluate LTSP seismic margins during modifications of certain plant components.

Ive included exerts of FSARU, Revision 20, in Appendix B.

PG&E implemented and maintained the CLB requirement for the SSE by the Plant Q-List. As shown in Appendix C, and required by 10 CFR 50, Appendix B; and the licensees commitment to Regulatory Guide 1.26,31 PG&E defined the facility SSE as the DDE in the facility design control management systems.32 30 The HE is not defined as part of the design bases. Per NEI 97.04, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, Appendix B, page B21, Seismic Topical Design Bases (ML003678532), design bases are explicitly established by regulatory requirements, primarily the GDCs. Since the HE is not tied to the GDCs or 10CFR50.55a, the HE is not part of the DCPP design bases. NEI-97.4 was endorsed by Regulatory Guide 1.186, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases. Maintaining selected plant SSCs qualified to the HE was a licensing bases commitment.

31 Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, required establishing quality classifications for those plant SSCs credited for preventing or mitigating design bases events as defined in the safety analysis.

32 Pacific Gas and Electric Company Nuclear Power Generation, Classification of Structures, Systems, and Components For Diablo Canyon Power Plant Units 1 And 2 (Q-LIST), Revision 27 NRC Form 690 Page 9

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Continued Item 11 In September 2013 (after the DPO was submitted), PG&E made extensive changes to FSARU Section 2.5, Geology and Seismology. Many of these changes affected the description of the seismic design basis. These changes also included addition of the Shoreline scenario as a lesser included case under the HE. PG&E did not screen these changes against the 10 CFR 50.59 criteria to determine if an amendment to the Operating License was required. PG&E justified omitting the required screen by stating these changed were derived from NRC correspondence:33 These enhancements are derived from correspondence with the NRC, NRC regulatory documentation, and specific USAFR text, therefore a 10 CFR 50.59 screen is not required.

Many of these changes indirectly addressed how SSC seismic safety functions were met. The 10 CFR 50.59 screening criteria required these changes to be evaluated:34 methods of evaluation included in the UFSAR to demonstrate that intended SSC design functions will be accomplished are considered part of the "facility as described in the UFSAR." Thus use of new or revised methods of evaluation is considered to be a change that is controlled by 10 CFR 50.59 and needs to be considered as part of this screening step. Changing elements of a method of evaluation included in the UFSAR, or use of an alternative method, must be evaluated under 10 CFR 50.59(c)(2)(viii) to determine if prior NRC approval is required. Changes to methods of evaluation (only) do not require evaluation against the first seven criteria.

These PG&E FSARU enhancements made to Section 2.5, Geology and Seismology may have contributed to the DPO Panels misunderstanding of the DCPP seismic design bases.

The Panel Assumed an Inappropriate Seismic Design Basis to Disposition the Issues Raised in the Differing Professional Opinion The Panel depositions of the DPO issues were based on the underlying assumption that both HE and DDE ground motions established the GDC 2 SSE design basis for the facility. Using this assumption, the Panel concluded that the higher of the two ground motions, either the DDE or the HE, established the bounding condition for seismic design. The Panel used this logic to conclude that an amendment to the Operating License was not required because the new seismic information was already bound by the HE ground motion.

For the Panels conclusions to be correct, then this underlying assumption must also be correct.

Unfortunately, the Panel Report did not include sufficient detail to provide the reader an understanding of how the Panel formed this understanding of the facility design bases.

In June 2014, I met with the Panel members. At the meeting, I stated that the CLB presented in the Panel Report appeared to be conflict with the FSARU (see Appendix B) and the DPO. I requested that the Panel provide the bases for this underlying CLB assumptions used to disposition the DPO. The Panel Chairman stated that the FSARU clearly established the HE as part of the facility design bases and he referred me to FSARU (Revision 21) Section 2.5.5.9, 33 DCPP Form 69-20108, UFSAR Change Request Section(s): 2. 5 (Seismology and Geology), June 2013 These enhancements are derived from correspondence with the NRC, NRC Regulatory documentation and specific UFSAR test, therefore a 10 CFR 50.59 screen is not required.

34 NEI 96-07, Guidelines for10 CFR 50.59 Evaluations (ML003636043), Section 4.2.1.3, Screening Changes to UFSAR Methods of Evaluation, as endorsed by Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, (ML003759710)

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Continued Item 11 Earthquake Design Basis. Ive included this FSARU Section below with highlighted changes incorporated with Revision 21 and PG&Es annotations (September 2013).35 A comparison of this FSARU Section with page A-6 (Appendix A), shows that PG&E added the HE as part of the seismic design bases description subsequent to plant licensing. This addition to the design basis description could be considered an acceptable change. However, the Panels use of this change to exclude the SSE/DDE requirements would be considered a change to the facility design bases and would require an amendment to the Operating License. 10 CFR 50.59 stated that an amendment to the Operating License was required before the licensee made a changed that result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.36 Consistent with the licensees commitment to Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (LWR Edition), FSARU Sections 3.1, Conformance with GDC, and 3.2.1, Seismic Classification, established the seismic design basis:

This section should identify those structures, systems, and components important to safety that are designed to withstand the effects of a Safe Shutdown Earthquake (see Section 2.5) and remain functional. These plant features are those necessary to ensure:

1. The integrity of the reactor coolant pressure boundary,

2. The capability to shut down the reactor and maintain it in a safe condition, or

3. The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100.

As shown in Appendixes A, B and C, the SSE for DCPP has always been the DDE, not the HE as described in the Panel Report..

The Panels assumption that the HE was included in the SSE design basis provided insufficient justification to exclude comparison of the new information against the DDE/SSE safety analysis. If both analyses supported the facility SSE, as described in the Panel Report, then both analyses must be required for GDC 2 compliance. If both analyses are required for GDC 2, then the bounding condition for comparison would include the DDE and the HE, not the Panels position of the DDE or the HE.

35 DCPP Form 69-20108, UFSAR Change Request Section(s): 2. 5 (Seismology and Geology), June 2013 36 For additional detail see: Nuclear Energy Institute, Guidelines For 10 CFR 50.59 Evaluations, February 22, 2000, Section 4.3.8, Does the Activity Result in a Departure from a Method of Evaluation Described in the UFSAR Used in Establishing the Design Bases or in the Safety Analyses?

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Continued Item 11 For the purposes the DPO disposition, it makes no difference whether or not the HE was or was not part of the GDC 2 design bases. The effect of the new information on the DDE/SSE licensing requirements and operability would still require disposition in terms of the license and operability.

As discussed in the DPO, the DDE/SSE was more limiting for SSC seismic qualification than the HE. Given the 70-percent increase represented by the new ground motions, the limitations of the DDE/SSE safety analysis became even more pronounced.

The Panel Report Failed to Address the Specific Regulatory and Statutory Requirements Cited in the Differing Professional Opinion The DPO identified the regulatory framework and specific statutory requirements that the agency failed to enforce at DCPP. Many of these requirements were related to the facility as described in the Final Safety Analysis Report Update. The Panel Report did not include adequate detail for the reader to conclude that these requirements were satisfied.

The DPO Panel Report stated that an FSARU change was likely not required at all, let alone, something that required a license amendment.

However, Title 10 CFR 50.71(e) required the FSARU GDC 2 safety analysis to be updated:

FASR originally submitted as part of the application for the operating license, to assure that the information included in the FSAR contains the latest material developed.

The updated dated FSAR shall be revised to include the effects of all changes made in the facility or procedures as described in the FSAR; all safety evaluations performed by the licensee.. and all analysis of new safety issues performed Title 10 CFR 50.34(b) required the FSAR to include a safety analysis demonstrating that the GDC 2 design basis was satisfied:

The FSAR shall include information that described the facility, presented the design bases and limits on its operation, and presents the safety analyses of the SSCs and of the facility as a whole.

The Diablo Canyon license application (original FSAR, Amendment 85) included a safety analysis that demonstrated the GDC 2 and Part 100, Appendix A, SSE design basis was satisfied. This analysis included an evaluation of all earthquake faults within 75 miles of the site (with exception of the Hosgri fault). From this evaluation, this safety analysis developed a ground motion. The licensee used this ground motion as the design bases controlling parameter 37 to determine the amount of seismic stress plant SSCs would be exposed to following the DDE/SSE. The safety analysis, consistent with 10 CFR 50.34(b), included a description demonstrating that the functional design bases requirements of GDC 2 and Part 100, Appendix A, were meet for the SSCs listed in Regulatory Guide 1.29.38 37 The DPO included a detained description of how this design bases controlling parameter was developed and used for SSC seismic qualification, consistent with NEI 97.04, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, Appendix B, for providing examples and guidance acceptable to the staff for providing a clearer understanding of what constitutes design bases information.

38 Per 10 CFR 100, App A, III(c) and 10 CFR 50.34(a)(3))

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Continued Item 11 The licensees new seismic information concluded that the existing design bases controlling parameter (ground motion) as described in the FSARU safety analysis, could be exceeded. PG&E was required to update the FSARU with this new information because the bounds of the safety analysis were challenged, calling into question the conclusion that the GDC 2 functional requirements were still satisfied. The new information raised the question if the plant SSCs, required by the design bases to remain functional for the DDE/SSE, would remain seismically qualified at the higher ground motions, within the context of the existing safety analysis.

The failure of PG&E to take prompt corrective action(s) to restore the bounds of safety analysis and plant SSCs to regulatory requirements and the design bases39 was a violation of 10 CFR 50, Appendix B. Appendix B stated:

Criterion III, Design Control, required that applicable regulatory requirements and the design basis (50.2) and as specified in the license application (FSAR), for those SSCs to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions.

Criterion XVI, Corrective Actions, required that conditions adverse to quality, such as failures, nonconformances, are promptly identified and corrected.

The new information resulted in the design basis (as specified in the license application for GDC 2) to be no longer correctly translated in the specifications, drawings, procedures, and instructions. The new seismic information rendered the FSARU SSE safety analysis non-conforming with GDC 2. 10 CFR 50.71(e) ensures that fidelity is maintained between new information, the FSARU safety analysis, and the GDC functional requirements establishing the design bases.40 The HE was unaffected by the new information for two independent reasons:

1) The CLB (FSARU) stated that the HE only applied to an earthquake on the Hosgri fault, and the new information was not related to the Hosgri fault, and

2) The HE was not used to establish the plant GDC 2 seismic design basis. The HE safety evaluation was not included in the FSARU. A 10 CFR 50.34 safety evaluation was not required to be included in the FSARU because the HE was not used to demonstrate that design bases or design basis functional requirements (GDC) were met.41 FSARU Change Required a License Amendment The Panel Report did not address the specific issues identified in the DPO related to the failure of the licensee to obtain an amendment to the license supporting the required FSARU changes per 10 CFR 50.71(e). As an alternative, the Panel addressed the actual changes the licensee made to 39 GDC 2 and Part 100, Appendix A, functional design based required: 1) integrity of the reactor coolant pressure boundary, 2) capability to shut down the reactor and maintain it in a safe condition, and 3) the SSCs needed to prevent or mitigate the consequences of accidents would remain functional given the maximum earthquake potential based on local geology.

40 10 CFR 50.71, Maintenance of Records, Making of Reports, implemented by Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e), ML003740112, and Section 5 of NEI 98-03, Revision 1, Guidelines For Updating Final Safety Analysis. Changes to the FSAR may only be made after the licensee demonstrates that an amendment to the Operating Licensee is not required per 10 CFR 50.59.

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Continued Item 11 the FSARU, Revision 21. The Report stated: Consequently, there was insufficient basis to conclude that a license amendment was required to address the 2011 Shoreline report, and the NRC staffs recommendation for an FSAR updated was reasonable.

FSARU changes per 10 CFR 50.71(e), are subject to the previsions of 10 CFR 50.59.42 10 CFR 50.59 stated:

A licensee shall obtain a license amendment pursuant to 50.90 prior to implementing a change, test or experiment if the change test or, experiment would:

- Results in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety, or

- Results in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analysis The new seismic information directly affected the information used in the FSARU safety analysis demonstrating that the GDC 2 design basis was satisfied. The licensee considered two cases.

For the first case, the licensee may update the existing FSARU safety analysis with the higher ground motions represented by the new seismic information. This update would result in the analyzed seismic stress to exceed ASME Code acceptance limits for reactor coolant system pressure boundary, major structures (reactor containment and auxiliary building), and the established qualification limits for important to safety SSCs (Regulatory Guide 1.29). NEI 96-0743 (Section 4.3.2) stated that a change to the facility as described in the FSARU that results in exceeding limits for seismic qualification required prior NRC approval because of the increased likelihood of a malfunction of SSCs important to safety (during an earthquake).

For the second case, the licensee may use a different analytical method to demonstrate that the GDC 2 design basis was still satisfied given the increased ground motions. The licensee determined that HE methodology could be applied to the new ground motions without exceeding established plant SSC seismic qualification limits. This case also required prior NRC approval because the new or proposed method (the HE) yielded results that were non-conservative when compared to the FSARU method (NEI 96-07, Section 4.3.8).

As required by 10 CFR 50.59 and 10 CFR 50.90, the licensee requested NRC approval to use the HE method (LAR 2011-05) to demonstrate that the GDC 2 design basis was satisfied at the higher ground motions. The NRC subsequently concluded that the HE method was not appropriate for the SSE and requested that the licensee withdrawn the LAR.

Similarly, the licensees action to revise the FSARU (Revision 21) to include the Shoreline (and presumably the San Luis Bay and Los Osos) fault(s) as lessor case(s) of the HE also required prior NRC approval. All of these faults are physically located within 75 miles of the site and are not 42 Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e),

ML003740112, and NEI 98-03, Revision 1,Guidelines For Updating Final Safety Analysis. Changes to the FSAR may only be made after the licensee demonstrates that an amendment to the Operating Licensee is not required per 10 CFR 50.59.

43 Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, (ML003759710) endorsed NEI 96-07, Guidelines for10 CFR 50.59 Evaluations ML003636043) as an acceptable method for implementation of 10 CFR 50.59.

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Continued Item 11 associated with the Hosgri fault. As defined in the CLB (FSARU Section 2.5), deterministic ground motions that may be produced by these faults are within the scope of the GDC 2 SSE safety analysis. To limit the effect of these new faults on plant SSC to only the HE methodology was also a change to the facility as described in the FSARU. The end result was to exclude the Shoreline, San Luis Bay, and Los Osos faults from the GDC 2 design basis and safety analysis. This action also required prior NRC approval because the new or proposed method (the HE method) yielded results that were non-conservative when compared to the FSARU method (NEI 96-07, Section 4.3.8).

Technical Speciation Operability The Panel Report stated:

For situations without specific technical specification testing requirements, evaluations can be performed by the licensee to determine if the equipment can still perform its design function using appropriate evaluation methods. There is not a regulation that requires the methods used in the original design calculations must be used in these evaluations. Many times, engineering evaluation methods have changed since the original Construction Permit application was made. This is particularly true for seismic hazards. Modern methods are frequently used to show the equipment can still perform its function. Typical equipment installed at the facility had margin above the minimums that the design basis calculations required.

The Panel concluded that NRC operability guidance (IMC 0326)44 allowed the licensee to use an alternative method for demonstrating that SSC specified safety functions could still be met at the higher ground motions. The Panel Report stated that the use of the HE or LTSP is attractive because the methods used in the LTSP are improved over those of initial licensing.

The Panel Report did not address the specific issues raised in the DPO related to the licensees use of these alternative methods. The DPO stated that the licensees use of the HE (or the LTSP) was inappropriate for operability because these methods over-predict SSC performance when compared to the GDC 2 CLB analysis methods. The NRC provides use of alternative methods45 to allow latitude for complex operability evaluations. The NRC restricts use of alternative methods that create additional margin when compared to the design basis method. For the new seismic information, the licensee had already established that SSC acceptance limits were exceeded using the GDC 2 design basis method. At this point, the licensee should have declared these SSCs inoperable and applied the required technical specification actions.

The DPO stated that the ASME Code acceptance limits are exceeded for reactor coolant pressure boundary components when the SSE seismic stresses are adjusted for the new higher ground motions. The Panel Report stated:

The FSARU identifies both the DDE and the Hosgri as faulted conditions for use in the seismic stress levels for appropriate component and piping and demonstrates how it meets the appropriate ASME acceptance criteria.

The use of both the DDE and the Hosgri in the evaluation is consistent with Panels conclusion that both these limits are, at times, applicable as the limiting load.

44 Inspection Manual Chapter 0326, Operability Determinations and Functionality Assessments for Conditions Adverse to Quality or Safety (ML13274A578) 45 (IMC 0326, Appendix C-04)

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Continued Item 11 The Panel conclusion was based on the assumption that either the HE or SSE methodology could be used to satisfy Code requirements. Since the new ground motions were lower than those assumed for the HE, the HE method would result in meeting Code acceptance limits (assuming that the licensee included the required load combinations).

The Panels conclusion did not consider the specific ASME Code and CLB requirements. The CLB, the Code, and 10 CFR 50.55a required the licensee to demonstrate that combined accident and SSE seismic loading be maintained below acceptance limits. Calculating the HE loading alone did not satisfy this requirement. The CLB clearly established the DDE as the SSE46. The HE was not the SSE. Neither the Code nor NRC Operability policy included provision to substitute the HE for the DDE/SSE to satisfy Code compliance. As a minimum, the DDE/SSE loads must meet acceptance limits. Also, as described in the DPO, for a given ground motion, the calculated stress will always be more limiting for the DDE/SSE method than for the HE. Because the Code specified that SSE loads be used, an amendment to the Operating License modifying the facility SSE design bases would be required before the HE could be used for Code compliance.

As described in the DPO, Code limits are exceeded when applying the new ground motions to the existing SSE Code calculations. Contrary to the Panel Report, IMC 0326, Appendix C.11, stated that a responsible expectation of operability cannot exist when Code requirements are not satisfied:

ASME Class 147 components do not meet ASME Code or construction code acceptance standards, the requirements of an NRC endorsed ASME Code Case, or an NRC approved alternative, then an immediate operability determination cannot conclude a reasonable expectation of operability exists and the components are inoperable. Satisfaction of Code acceptance standards is the minimum necessary for operability of Class 1 pressure boundary components because of the importance of the safety function being performed.

PG&E should have immediately declared ASME Class 1 components (reactor coolant pressure boundary) inoperable once they concluded exceedances existed with the higher ground motions.

The CLB stated that licensee demonstrated that Code limits were met for certain HE faulted cases. However, neither the ASME Code nor 10 CFR 50.55a required the licensee to perform these calculations. The license performed these calculations to meet a licensing bases commitment, not to satisfy design bases or a regulatory requirement.

Existing NRC Expectations Following Discovery of New Conditions Outside the Bounds of the Safety Analysis The DPO Panel Report transmittal letter stated:

Finally, the Panel concluded that the lack of formal regulatory guidance for evaluating new information of natural hazards appears to be a contributing cause in creating many of the differing interpretations for potential significance of the information, along with confusion with regard to the regulatory process for evaluating the impact of new seismic information on system operability.

The agency has provided sufficient formal regulatory guidance for evaluating new information, including information affecting natural hazards. The DPO was written because the NRC staff failed 46 See Appendix A and B of this report. DDE is the SSE for DCPP and HE did not include accident LOCA loads.

47 Class 1 components make up the reactor coolant pressure boundary and pipe/component supports.

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Continued Item 11 to follow this formal guidance during disposition of the Diablo Canyon seismic issues. This existing guidance included:

1) NRC Regulatory Issues Summary (RIS) 2013-05:48 This RIS addressed questions raised about the relationship between licensing basis design requirements, the GDCs, and technical specification operability.

It is the staffs position that failure to meet a GDC, as described in the licensing basis (e.g., non-conforming with the CLB for protection against flooding, seismic, tornadoes) should be treated as a nonconforming condition and is an entry point for an operability determination if the non-conforming condition calls into question the ability of the SSCs to perform their specified safety functions(s) or necessary and related support functions(s).

The safety analysis report describes the design capability of the facility to meet the GDC (or a plant-specific equivalent). The staff safety evaluation report documents the acceptability of safety analysis report analyses.

The analyses and evaluation included in the safety analysis serve as the basis for TS issued with the operating license. The TS limiting conditions for operation, according to 10 CFR 50.36(c)(2)(i), are the lowest functional capability or performance levels of equipment required for safe operation of the facility. Section 182 of the Atomic Energy Act of 1954, as amended and as implemented by 10 CFR 50.36, requires that those design features of the facility that, if altered or modified, would have a significant effect on safety, be included in the TS.

Thus, TS are intended to ensure that the most safety significant design features of a plant, as determined by the safety analysis, maintain their capability to perform their safety functions, i.e., that SSCs are capable of performing their specified safety functions or necessary and related support functions.

Thus, an operability determination is appropriate upon identification of a degraded or nonconforming condition that calls into question the ability of SSCs to perform their specified safety function, including any nonconforming condition with a GDC included in either the CLB for an SSC described in TS or for a necessary and related support function required by the definition of operability. If the licensee determination concludes that the TS SSC is nonconforming but operable or the necessary and related support function is nonconforming but functional, it would be appropriate to address the nonconforming condition through the licensees corrective action program.

2) Formal NRC regulatory guidance letter related to seismic hazard reevaluations:49 This supplemental information reinforced agency regulations to address non-conforming conditions associated with the CLB:

During the course of stakeholder interactions regarding the hazard reevaluations, various questions were raised with respect to operability and reportability of systems, structures, and components (SSC) if the reevaluated seismic hazard is not bounded by the current seismic design basis.

However, as with any new information that may arise at a plant, licensees are responsible for evaluating and making determinations related to operability, and any associated reportability, on a case-by-case basis.

Licensees should consider and disposition the information through their corrective action program or equivalent process. If an error is identified in the current design or licensing basis during the performance of the requested seismic hazard evaluation, the staff expects that licensees would assess the operability of the affected SSC. Additionally, licensees would need to determine if the situation is reportable pursuant to 10 CFR 50.72 and 50.73. Licensees would also be expected to determine whether aspects of 10 CFR 50.9, concerning the requirement to provide complete and accurate information to the NRC, would be applicable.

48 RIS 2013-05, NRC Position on the Relationship between General Design Criteria and Technical Specification Operability (ML13056A077) 49 Letter from E Leeds, Supplemental Information Related To Request For Information Pursuant To Title 10 of The Code Of Federal Regulations 50.54(f) Regarding Seismic Hazard Reevaluations For Recommendation 2.1 of the Near-Term Task Force Review of Insights From The Fukushima Dai-Ichi Accident, February 20, 2014 (ML14030A046)

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Continued Item 11 At DCPP, PG&E developed new information that identified invalid inputs (errors) were used in the CLB safety analysis that demonstrated that the GDC 2 seismic design basis was met.

3) Inspection Manual Chapter 0326:50 IMC provided formal regulatory guidance for evaluating new information of natural hazards. Section C.1 stated:

Failure to meet GDC, as described in the licensing basis (e.g., nonconformance with the CLB for protection against flooding, seismic events, tornadoes) should be treated as a nonconforming condition and is an entry point for an operability determination if the nonconforming condition calls into question the ability of SSCs to perform their specified safety function(s) or necessary and related support function(s). If the licensee determination concludes that the TS SSC is nonconforming but operable or the necessary and related support function is nonconforming but functional, it would be appropriate to address the nonconforming condition through the licensees corrective action program. However, if the licensees evaluation concludes that the TS SSC is inoperable, then the licensee must enter its TS and follow the applicable required actions.

4) The NRC enforced CLB GDC 2 flooding requirements at Watts Bar.51 Tennessee Valley Authority personnel identified that the spillway coefficient used to model flow from an upstream dam needed to be updated. Utility engineers found that the updated coefficient reduced the amount of spillway flow expected during periods of heavy rain. The reduction of spillway flow affected safety analysis inputs used to demonstrate that the facility met the GDC 2 design bases for maximum flood height. This case was very similar to the DCPP. At both facilities, new information affected the outcome of GDC 2 safety analyses and the capability of plant SSCs to perform the required safety functions. In the Watts Bar case, the new information resulted in a higher maximum flood height. In the DCPP case, the new information resulted in an increase in the amount of seismic stress affecting plant SSCs following an earthquake. In both cases, the licensees failed to take prompt corrective actions to correct the non-conforming safety analysis. However, for the Watts Bar case, the agency enforced statutory design control requirements. This enforcement action included:

- A Severity Level III violation for failing to report an unanalyzed condition related to external flooding

- A Yellow Finding following the failure to maintain an adequate abnormal condition procedure to implement the flood mitigation strategy

- A White Finding following inadequate abnormal condition procedure for flood mitigation strategy.

5) The NRC also enforced GDC 2 CLB flooding requirements at several other facilities. For example, the NRC issued a Yellow Finding at the Monticello facility.52 In the Monticello case, the licensee was unable to implement flood protection barriers consistent with the GDC 2 flooding safety analysis.

50 IMC 0236, Operability Determinations and Functionality Assessments for Conditions Adverse to Quality or Safety (ML13274A578), Section 3.60 defined nonconforming condition and Section C-1 included the failure to meet a GDC as a non-conforming condition, Section C-11 defined the requirement to meet ASME 51 Watts Bar Unit 1 Nuclear Plant - Final Significance Determination Of Yellow Finding, White Finding And Notices Of Violations; Assessment Follow-Up Letter; Inspection Report No. 05000390/2013009, EA-13-018, June 4, 2013.

52 Final Significance Determination of A Yellow Finding With Assessment Follow up and Notice of Violation; NRC Inspection Report No. 5000263/2013009; Monticello Nuclear Generating Plant, EA-13-096, August 28, 2013.

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Continued Item 11 Fukushima Term Task Force Recommendations 2.1 and 2.3 The Panel Report and Research Information Letter 12-0153 both stated that the Fukushima Recommendation 2.1, Seismic Reevaluations,54 will address the DCPP seismic issues. While the seismic reevaluations are designed to assess the seismic hazard for the facility, these ongoing activities do not address the concerns raised in the DPO. The DPO focused on the failure of agency personnel in enforce CLB requirements, not on how seismic hazards are evaluated. The requested seismic reevaluation will provide context for the agency to determine if the CLB should be modified.

In contrast, one purpose of Recommendation 2.3,55 was to confirm that CLB seismic requirements were met while the seismic reevaluations are performed. Verification that the plant was operating within the bounds of the current design and licensing bases provided confidence that the plant was safe while the reevaluations are performed:

Structures, systems, and components (SSCs) important to safety in operating nuclear power plants are designed either in accordance with, or meet the intent of, Appendix A to 10 CFR Part 100 and Appendix A to 10 CFR Part 5O, General Design Criteria (GGC) 2. GDC 2 states that SSCs important to safety at nuclear power plants must be designed to withstand the effects of natural phenomena such as earthquakes, tornados, hurricanes, floods, tsunami, and seiches without loss of capability to perform their intended safety functions.

The design bases for these SSCs are to reflect appropriate consideration of the most severe of the natural phenomena that have been historically reported for the site and surrounding area. The design bases are also to reflect sufficient margin to account for the limited accuracy, quantity, and period of time in which the historical data have been accumulated.

In response to NTTF Recommendation 2.3, the Commission requests all licensees to perform seismic walkdowns in order to identify and address plant specific degraded, nonconforming, or unanalyzed conditions and verify the adequacy of strategies, monitoring, and maintenance programs such that the nuclear power plant can respond to external events. The walkdown will verify current plant configuration with the current licensing basis, verify the adequacy of current strategies, maintenance plans, and identify degraded, nonconforming, or unanalyzed conditions.

If any condition identified during the walkdown activities represents a degraded, nonconforming, or unanalyzed condition (i.e., noncompliance with the current licensing basis) for an SSC, describe actions that were taken or are planned to address the condition using the guidance in Regulatory Issues Summary 2005-20, Revision 1, Revision to NRC Inspection Manual Part 9900 Technical Guidance, "Operability Conditions Adverse to Quality or Safety," including entering the condition in the corrective action program. Reporting requirements pursuant to 10 CFR 50.72 should also be considered. Additionally, these findings should be considered in the Recommendation 2.1 hazard evaluations, as appropriate.

As detailed in the DPO, DCPP continues to operate in both unanalyzed and non-conforming conditions outside of the bounds of the CLB.

53 Diablo Canyon Power Plant, Unit Nos. 1 And 2 -NRC Review of Shoreline Fault (TAC NOS. ME5306 AND ME5307),

October 12, 2012 (ML120730106).

54 Request For Information Pursuant To Title 10 Of The Code of Federal Regulations 50.54(F) Regarding Recommendations 2.1,2.3, And 9.3, of The Near-Term Task Force Review Of Insights From The Fukushima Dai-Ichi Accident (ML12053A340) 55 See Footnote 51.

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Continued Item 11 Summary The existing regulatory framework for addressing the enforcement and operability issues raised in DPO 2013-002 are well established. NRC regulations56 required PG&E to take prompt corrective action after developing new seismic information that concluded that the GDC 2 safety analysis was no longer bounding for the seismic qualification of plant SSCs. These actions also required the licensee to either incorporate the new seismic information into the existing safety analysis or establish a new methodology for demonstrating that the functional design bases requirements of GDC 2 remained satisfied.57 Either approach required an amendment to the DCPP Operating License per 10 CFR 50.5958 and 10 CFR 50.90.

PG&E requested that the NRC approve the HE, as a new method for the facility SSE. However, the NRC concluded that this new methodology was not appropriate for establishing the facility SSE and requested that the licensee withdraw the LAR. After the license amendment process was unsuccessful, the NRR PM provided the licensee direction to work around the amendment process by directly adding the new information to the FSARU. This action subverted the license amendment public notice requirements and hearing opportunities as prescribed by 10 CFR 50.91.

PG&E continued to operate the DCPP reactors following discovery of the unanalyzed condition and non-conforming safety analysis. The licensee was required to demonstrate that technical specifications SSCs would still be capable of performing the safety functions specified in the safety analysis at the higher seismic stress levels. The licensees use of the HE alternative method for this demonstration was not consistent with NRC policy. The HE was inappropriate because for a given ground motion, the HE would always over-predict SSC seismic performance when compared to the SSE design basis method. Also, the licensees use of the HE to demonstrate that reactor coolant pressure boundary integrity would be maintained during an earthquake was inconsistent with ASME Code requirements and 10 CFR 50.55a.

The DPO Panel concluded that an amendment to Operating License was not required to disposition the new seismic information. The Panel also concluded that the licensee satisfied all statutory requirements. The Panels conclusions were based on the inappropriate assumption that GDC 2 SSE design basis was established by a combination of the DDE safety analysis and the HE. From this assumption, the Panel extrapolated that the new information was within the existing SSE GDC 2 design basis because the new ground motions were bound by either the DDE or the HE. The Panel Report did not include the bases for either of these assumptions.

This DPO Appeal demonstrates that the Panels conclusions were incorrect because the underlying assumptions used to formulate those conclusions were inconsistent with the CLB. The CLB clearly described that the DDE was the facility SSE and the supporting DDE safety analysis demonstrated that the GDC 2 design basis was met. Even if the HE was considered part of the 10 CFR 50.2 design bases, then Panel Report provided inadequate justification to exclude the 56 Appendix B to Part 50, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Criterion III. Design Control, and XVI. Corrective Action.

57 10 CFR50.71(e) required the FSARU to include all analyses of new safety issues affecting the originally license application to assure that the information included in the report contains the latest information developed 58 10 CFR 50.59 required an amendment to the Operating License for FSARU changes that result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses.

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Continued Item 11 DDE/SSE safety analysis from the requirements of 10 CFR 50.59, 50.71(e), and Part 50, Appendix B. In either case, the new ground motions must be evaluated within the context of GDC 2 design bases and limiting SSC seismic qualification requirements.

Requested Action Please take the following actions:

1. Disapprove the Panel Report depositing DPO 2013-002.

2. Initiate regulatory enforcement action to address the ongoing non-compliances with Part 50, Appendix B, 10 CFR 50.59, and plant technical specifications at DCPP.

3. Initiate a review to determine why the non-concurrence (NCP 2012-01) and the DPO process were not effective to address the outstanding DCPP seismic issues.

Thank you, Michael Peck, Ph.D.

Attachments:

Appendix A, Original Diablo Canyon Seismic Licensing Bases Appendix B, Current Diablo Canyon Seismic Licensing Bases Appendix C, Pacific Gas and Electric Company Nuclear Power Generation, Classification of Structures, Systems, and Components for Diablo Canyon Power Plant Units 1 And 2 (Q-LIST),

Revision 27 NRC Form 690 Page 21

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