ML14252A743

DPO Case File-DPO-2013-002 (Public)

ML14252A743
Person / Time
Site:	Diablo Canyon
Issue date:	09/09/2014
From:	Sewell M NRC/OE
To:
Sewell M
References
DPO-2013-002, FOIA/PA-2014-0488
Download: ML14252A743 (164)

Text

DPO Case File for DPO-2013-002 The following pdf represents a collection of documents associated with the submittal and disposition of a differing professional opinion (DPO) from an NRC employee involving seismic issues at Diablo Canyon.

Management Directive (MD) 10.159, The NRC Differing Professional Opinions Program, dated May 16, 2004, describes the DPO Program.

http://pbadupws.nrc.gov/docs/ML0417/ML041770431.pdf The DPO Program is a formal process that allows employees and NRC contractors to have their differing views on established, mission-related issues considered by the highest level managers in their organizations, i.e., Office Directors and Regional Administrators. The process also provides managers with an independent, three-person review of the issue (one person chosen by the employee). After a decision is issued to an employee, he or she may appeal the decision to the Executive Director for Operations (EDO).

Because the disposition of a DPO represents a multi-step process, readers should view the records as a collection. In other words, reading a document in isolation will not provide the correct context for how this issue was considered by the NRC.

The records in this collection have been reviewed and approved for public dissemination.

Document 1: DPO Submittal Document 2: Memo from Office Director Establishing DPO Panel Document 3: DPO Panel Report Document 4: DPO Decision Document 5: DPO Appeal Submittal Document 6: Office Directors Statement of Views Document 7: DPO Submitters Appeal Presentation to OEDO Document 8: DPO Appeal Decision

Document 1 - DPO Submittal NRC FORM 680 U.S. NUCLEAR REGULA TORY COMMISSION FOR PROCESSING USE ONLY (11.2002)

NRCMD 10.159

1. DPO CASE NUMBER .

DIFFERING PROFESSIONAL OPINION lJ ~() -;)_o/3 ~ ()<)"<

INSTRUCTIONS: Prepare this form legibly and submit three copies to tile address provided in Block 14 below. ~A7i;EIVEi ~0 t 3

3. NAME OF SUBMITIER 4. POSITION TITLE v . GRADE Senior Reactor Instructor GG-14 r*

Michael Peck

6. OFFIC£/OIVISIONIBRANCHISECnON 7. BUILDING 9 . SUPERVISOR MAJLSTOP OCHCO/ADHRTD/RTTBB nc Steve Rutledge,

10. PRESENT SITUATION, CONDITION. METHOD, ETC., WHICH YOU BELIEVE SHOULD BE CHANGED OR IMPROVED.

(ConUnue on Page 2 or 3 as necessary.)

Please see attachment for the DPO.

Please note: This DPO involves Region IV, Reactor Projects, and NRR, DORL. These issues were developed while I was the senior resident inspector at Diablo Canyon. My supervisor was Neil O'keefe.

I was subsequently reassigned to the TTC.

11. DESCRIBE YOUR DIFFERING OPINION IN ACCORDANCE V>.rrH Tl1E GUIDANCE PRESENTED IN NRC MANAGEMENT DIRECTIVE 10.159.

(Continue on Page 2 or 3 as necessary.)

Please see attachment.

As discussed in MD 10.159, please make this DPO available to the public.

Thank you,

12. Checl< (a) or (b) as appropriat&:

[{]a. Thorough discussions of the issue{s) raised in item 11 have taken place within my management chain; or Db. The reasons why I cannot approach my immediate chain of command are:

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13. PROPOSED PANEL MEMBERS ARE (in prlon'tyordetj:

rr:u b .

I r;J..ol;;

SIGNATURE Of CO-SUBMITIER

14. Submit this form IP:

(if any)

'DATE

1. Gerond George Differing Professional Opinions Program Manager y***y********y * *

2. Larry Criscione Office or: OE/CRB

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3. Rudy Bernhard Mail Stop: 4A15A

15. ACKNOWLEDGMENT K: OFO!FF71NG~ANAGER (DPOPM)

THANK YOU FOR YOUR DIFFERING PROFESSIONAL OPINION. It will be carefully considered by a panel o I c ~*".. '

experts in accordance with the provisions of NRCMD PR.E..C,<JNDITIONS MET D~ or ;O; 10.159. and you will be advised of any action taken. Your interest in improving NRC operations is appreciated. {i_YES D NO E{ ; :T I3 NRC FORM 680 (11-2002\

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Differing Professional Opinion - Diablo Canyon Seismic Issues 1.0 Summary In 2011, Pacific Gas and Electric (PG&E) submitted a report to the NRC that included a reevaluation of the local geology surrounding the Diablo Canyon Power Plant. 1 This report included deterministic evaluations concluding that three local earthquake faults are capable of generating significantly greater vibratory ground motion than was used to establish the facility safe shutdown earthquake (SSE) design basis. In response to this issue, NRC staff actions have been inconsistent with existing regulatory requirements and the facility design bases and Operating License.

a. Less than Adequate Corrective Actions to Incorporation the New Seismic Information Into the Current Licensing Basis (CLB)

Prevailing Staff View: The NRC concluded that potential earthquake ground motions from the Shoreline fault are at or below those levels for which the plant was previously evaluated and demonstrated to have a reasonable assurance of safety. 2 The staff stated that PG&E should incorporate Shoreline scenario into the Final Safety Analysis Report Update (FSARU) as an included case under the Hosgri evaluation (HE).

Alternate View: Incorporating the Shoreline scenario into the FSARU will require an amendment to the Diablo Canyon Operating License. A license amendment is required because the change results in more than a minimal increase in the likelihood of a malfunction of a structure, system, or component (SSC) important to safety than previously evaluated in the FSARU. A license amendment is also required because this change represents a departure from the FSARU method of evaluation used to establish the seismic SSE design basis. PG&E previously submitted a license amendment request to modify the plant design bases and safety analysis to accommodate the new seismic information. However, this request was not accepted by the NRC for review.

The staffs conclusion of a reasonable assurance of safety does not provide an acceptable basis for not enforcing existing NRC quality assurance, safety analysis, and license requirements. The staffs corrective action also failed to address the Los Osos and San Luis Bay faults. The new seismic information concluded that these faults were also capable of producing ground motions in excess of the current plant SSE design basis.

Recommended Action: The NRC to initiate enforcement action to ensure PG&E complies with NRC quality assurance requirements to take prompt corrective action to correct the nonconforming FSARU safety analysis.

b. Failure to Demonstrate Plant Technical Specification Required Structures, Systems, and Components (SSCs) are Operable Prevailing Staff View: The NRC concluded that all Diablo Canyon technical specification required plant SSCs were operable at the higher ground motions. 3,4 The staff based this conclusion on a comparison of the new seismic information with the ground motion spectrums used in the HE and the Long Term Seismic Program (LTSP). 5 While the new ground motions exceeded those used to establish the SSE design basis and the NRC approved safety analysis, they were bound by the HE and LTSP.

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Alternate View: The prevailing staff view is contrary to the NRC operability policy. To be considered operable, a reasonable assurance must be demonstrated that nonconforming SSC are capable of the performing the safety function(s) specified by the design and within the required range of design physical conditions defined in the CLB, including the design bases. Neither the HE nor the LSTP contain design bases limits, conditions, or assumptions used in the bounding SSE safety analysis. Comparison of the new ground motions only against the HE and LSTP failed to demonstrate that all plant technical specification required SSCs are capable of meeting the specified safety functions established at the higher ground motions:

• Neither the HE nor the LTSP methods are approved for use in the Diablo Canyon SSE design basis or safety analysis. The CLB defined the HE as an exception to the SSE and was only approved for evaluating the Hosgri fault. The LTSP is not part of the seismic design basis or safety analysis.

• Use of the HE and LTSP over-predicts SSC performance when compared to the CLB SSE methods. Neither the HE nor the LTSP are bounding for SSC seismic qualification at Diablo Canyon. Comparisons limited to only ground motion are meaningless for operability. These comparisons omit other relative CLB requirements including the methods, assumptions, initial conditions, and acceptance criteria applicable to each evaluation.

• Comparison of the new information only to the HE and LTSP failed to demonstrate that the requirements of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code are met at the higher ground motions.

Operability requires that the Code acceptance criteria are met for key plant components, including the reactor coolant pressure boundary.

Recommended Action: The NRC to initiate enforcement action to ensure PG&E complies with plant technical specification required actions to shutdown the Diablo Canyon reactors. The reactors should remain shut down pending demonstration that SSC safety functions can be meet at the higher seismic stress levels or until the NRC approves necessary dispensation and/or exemptions from the applicable regulatory and Operating License requirements.

Assessment of the Consequences if submitters position is not adopted by the Agency: The new seismic information resulted in a condition outside of the bounds of the existing Diablo Canyon design basis and safety analysis. Continued reactor operation outside the bounds of the NRC approved safety analyses challenges the presumption of nuclear safety.

The prevailing staff view that operability may be demonstrated independent of existing facility design bases and safety analyses requirements establishes a new industry precedent. Power reactor licensees may apply this precedent to other nonconforming and unanalyzed conditions.

2.0 Introduction The Atomic Energy Act of 1954, as amended, establishes "adequate protection" as the standard of safety on which NRC regulation is based. In the context of NRC regulation, safety means avoiding undue risk or providing reasonable assurance of adequate protection 2

for the public. Safety is the fundamental regulatory objective, and compliance with NRC requirements plays a fundamental role in providing confidence that safety is maintained.

NRC requirements have been designed to ensure adequate protection, which in turn, corresponds to "no undue risk to public health and safety. This goal is met through acceptable design and quality assurance measures. In the context of risk-informed regulation, compliance plays a very important role in ensuring that key assumptions used in underlying risk and engineering analyses remain valid. 6 Adequate protection is presumptively assured by compliance with NRC requirements.

These requirements limit plant operation within the design bases. These regulations also required that licensees establish, maintain, and operate within the boundaries of the NRC approved safety analyses. Operation within the bounds of the safety analysis provides confidence that the plant response to accidents and events will be consistent with the design bases.

At Diablo Canyon, the licensee developed new information that revealed that an unforeseen hazard exists. This new information concluded that three local faults are capable of producing earthquakes greater than those bound by the Diablo Canyon safe shutdown earthquake (SSE) design basis. The presumption of nuclear safety is challenged because plant operation is no longer within the bounds of the design basis and quality assurance measures the NRC used to license the facility.

A nonconforming condition exists when the plant safety analysis no longer meets NRC design bases and regulatory requirements. An unanalyzed condition exists when reactor operation occurs outside of the limiting bounds established in the NRC approved safety analysis. The Diablo Canyon seismic information resulted in both nonconforming and unanalyzed conditions. NRC quality assurance requirements required PG&E to implement prompt corrective actions to either restore the plant configuration within the bounds of the safety analysis or request NRC approval to revise the plant Operating License to accommodate the new information. The NRC has not enforced these regulatory requirements to correct the deficient seismic safety analysis at Diablo Canyon.

The NRC staff has discussed Diablo Canyon seismic issues for the past several years.

Several staff members viewed the new PG&E seismic information as beyond the existing regulatory framework. These staff members proposed new regulatory processes to review and disposition this information. These recommendations were similar to those proposed for the resolution to Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, and provided by the Fukishima Near-Term Task Force. These approaches request licensees compare the results of newly developed probabilistic ground motions models against the existing deterministic SSE. Subsequent Regulatory decisions are made based on the risk insights gained from these comparisons.

The updated Diablo Canyon seismic information was unique because PG&E included detailed deterministic evaluations of the local geology. These deterministic evaluations provided a one-to-one correspondence to seismic evaluations included in the CLB.

Comparing this new information with the CLB indicated that the plant was operating outside the bounds of the existing safety analysis. This called into question if the plant design bases requirements could still be met following an earthquake. From an inspection point of view, the regulatory framework for addressing nonconforming safety analyses and unanalyzed conditions are familiar. The PG&E case was different because these conditions were 3

specifically related to the seismic design basis, an area rarely touched by the Inspection Program prior to the Fukishima accident.

The integrity of key assumptions used in the safety analyses are maintained by requiring licensees to comply with the plant technical specifications. Technical specifications require plant operators to implement time dependent actions, including shutting down the reactors, when prescribed SSCs are no longer operable. Following identification of nonconforming or unanalyzed conditions, the operability process provides assurance that the plant is safe to continue to operate during the corrective action period. To be considered operable, plant SSCs must be capable of performing the safety functions described in the CLB, including the FSARU safety analyses. These safety functions include the capability to prevent or mitigate accidents and events following the vibratory motion (shaking) associated with the SSE. The staff concluded that all Diablo Canyon SSCs were operable using an alternative basis. However, the operability process did not provide the staff the flexibility to use this alternate approach. While the NRC has statutory authority to amend the facility Operating License to allow use of these alternate bases or exempt PG&E from regulatory requirements, the staff did not implement either of these processes to waive the Diablo Canyon CLB requirements.

3.0 Diablo Canyon Current Licensing Basis (CLB)

NRC regulations use the terms safety analysis, design bases, and nonconforming condition within the context of the CLB. A clear understanding how the NRC defined these terms and the specific Diablo Canyon License requirements are needed before the seismic corrective actions and operability can be assessed. The CLB includes the set of NRC requirements applicable to nuclear power plant license plus the docketed and currently effective written commitments for ensuring compliance with these NRC requirements and the plant-specific design basis. 7 For Diablo Canyon, seismic CLB explicitly includes:

• NRC regulations in 10 CFR Parts 2, 50, 100 (including Appendixes)

• Plant-specific design basis information, as defined in 10 CFR 50.2, and documented FSARU as required by 10 CFR 34 and 50.71(e)

• Plant technical specifications Design Bases Title 10 of the Code of Federal Regulations, Part 50.2, defines design bases as that information which identified the specific functions to be performed by plant SSCs and the specific values or ranges of values chosen for controlling parameters as reference bounds for the design. The NRC endorsed an expanded definition of design bases in NEI 97-04, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, Appendix B. 8 This expanded definition of design bases included:

• Design Bases Functions: Functional requirements derived from the principal design criteria used for Diablo Canyon. These establish the minimum standards set by 10 CFR Part 50, Appendix A, General Design Criteria (GDC), and other NRC regulations imposing functional requirements or limits on the plant design. For plant SSCs, design bases function include those:

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(1) required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications, or (2) credited in licensee safety analyses to meet NRC requirements.

For seismic qualification, the design basis functional requirements are established by 10 CFR 50, GDC 2, and 10 CFR 100, Appendix A. 9

• Design Bases Values: Values or ranges of values used for the controlling parameters establishing the reference bounds for the design and to meet the design bases functional requirements. These values may be:

(1) established by NRC requirement, (2) derived from or confirmed by safety analyses, or (3) chosen by the licensee from an applicable code, standard or guidance document.

Design bases values include the bounding conditions under which SSCs must perform the design bases functions for normal operation or following accidents or events. Plant specified events include those specified in the regulations, including the SSE.

Design Bases Controlling Parameters: Values chosen as reference bounds for the design. For example, for the seismic design basis, the SSE ground motion spectra are a design bases controlling parameter. 10 The CLB also includes supporting design information. While supporting design information is not explicitly part of the design bases, this information includes assumptions and inputs used in the safety analysis and by the NRC to verify design basis acceptance limits are met.

For seismic qualification, examples of supporting design information include:

• Commitment to NRC Safety Guide 29 (Regulatory Guide 1.29), Seismic Design Classification. Safety Guide 29 provides an NRC approved list of plant SSCs that are required to be qualified for the SSE.

• Methods used in the safety analysis to establish the SSE response spectra.

• Seismic damping values used in the structural dynamic analysis The facility design bases are a subset of the CLB and are required to be included in the FSARU by 10 CFR 50.34 and 10 CFR 50.71(e).

Regulations Establishing the Seismic Design Bases Title 10 of the Code of Federal Regulations, Part 50, Appendix A, General Design Criteria (GDC) 2, 11 Design Bases for Protection against Natural Phenomena, established the design basis requirements for seismic qualification. SSCs important to safety must be capable of withstanding the effects of earthquakes without loss of capability to perform their safety functions. GDC 2 requires:

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• Appropriate consideration of the most severe natural phenomena that has been historically reported for the site and surrounding area, with sufficient margin for the limited accuracy, quantity, and period that historical data was accumulated;

• Appropriate combinations of the effects of normal and accident conditions with the effects of the natural phenomena; and

• The importance of the safety functions to be performed.

Title 10, Code of Federal Regulations, Part 100, Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, implements the GDC 2 requirements for seismic design.

SSCs important to safety must be capable of withstanding the effects of the SSE without loss of capability to perform their safety functions. Appendix A defines the SSE as the maximum earthquake potential considering the regional and local geology and seismology and specific characteristics of local subsurface material. Appendix A applies to those important to safety SSCs necessary to assure:

• The integrity of the reactor coolant pressure boundary,

• The capability to shut down the reactor and maintain it in a safe shutdown condition,

• The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures.

Safety Analysis: Demonstrates that the facility meets the design bases, the capability to withstand or respond to postulated events, and that NRC acceptance criteria are met: 12,13,14 Seismic Qualification Process Pacific Gas and Electric seismically qualified plant SSCs (listed in Table 1) that are required to remain functional following the SSE. The seismic qualification process was generally preformed in three steps:

a. Evaluation of the local geology (FSARU Section 2.5)

This evaluation examined the local geology and deterministically identified the maximum earthquake potential that could affect important to safety plant equipment.

The safety analysis used NRC approved ground motion and attenuation methods and assumptions to establish the maximum vibratory ground motion for the site. At Diablo Canyon, the maximum ground motion was called the double design earthquake (DDE) and is equivalent to the SSE defined in 10 CFR 100, Appendix A.

b. Attenuation of seismic energy to important to safety SSC (FSARU Section 3.7)

This evaluation established how much seismic energy, or shaking, each important to safety SSC would be exposed to following the SSE/DDE. The analysis used NRC approved attenuation models and design basis inputs to propagate the seismic energy through plant structures, equipment, and piping systems. These models and inputs are part of the facility CLB.

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c. SSC Seismic qualification (FSARU Sections 3.2, 3.8, 3.9, 3.10, & 5.2)

PG&E seismically qualified the plant SSCs listed in Table 1 to ensure they would remain functional at the level of shaking that was determined to occur at that plant location following the SSE/DDE. This qualification was performed by a combination of testing and analyses. The functionality of some plant SSCs were demonstrated by use of a shaker table test. Other SSCs were qualified by NRC approved analysis. For example, the reactor coolant pressure boundary, piping systems, and the containment structure were qualified by ensuring that the seismically induced stress would not exceed acceptance levels established by the ASME and other codes.

Table 1 - Plant SSCs Qualified to SSE/DDE Technical Diablo Canyon Plant Structures, and Systems Specification 15 Required to be Qualified to the SSE/DDE Required SSCs

1. The reactor coolant pressure boundary. Yes

2. The reactor core and reactor vessel internals. Yes

3. Systems required for

- Emergency core cooling system Yes

- Containment heat removal, Yes

- Shutdown the reactor shutdown, Yes

- Remove residual heat Yes

- Cooling the spent fuel storage pool, No

4. Steam and feedwater systems up to and including the outermost Yes containment isolation valves.

5. Cooling water that are required for:

- Emergency core cooling, Yes

- Post-accident containment heat removal Yes

- Residual heat removal from the reactor, or Yes

- Cooling the spent fuel storage pool. No

6. Cooling and seal water systems required for functioning of reactor coolant No system components important to safety (reactor coolant pumps).

7. Systems or portions of systems that are required to supply fuel for Yes emergency equipment.

8. All electric and mechanical devices and circuitry between the process and Yes the input terminals of the actuator systems involved in generating signals that initiate protective action

9. Systems or portions of systems required for monitoring of systems important Yes to safety and actuation of systems important to safety.

10. The spent fuel No

11. The spent fuel storage pool structure, including the fuel racks. No

12. The reactivity control systems, control rods, control rod drives and boron Yes injection system.

13. The control room, including its associated equipment and all equipment Yes needed to maintain the control room within safe habitability limits.

14. Primary and secondary reactor containment. Yes

15. Systems, other than radioactive waste management systems, (not covered No above) that contain or may contain radioactive material and whose postulated failure would result in conservatively calculated potential offsite doses (using approved dose methods).

16. The Class 1E electric systems, including the auxiliary systems for the onsite Yes electric power supplies, that provide the emergency electric power needed for functioning of plant engineered safety features.

17. Those portions of structures, systems, or components whose continued May affect TS function is not required but whose failure could reduce the functioning of any plant feature included above to an unacceptable safety level or could result in incapacitating injury to occupants of the control room should be designed and constructed so that the SSE would not cause such failure. Must meet

18. Seismic Category I design requirements should extend to the first seismic applicable Code restraint beyond the defined boundaries. requirements 7

Diablo Canyon FSARU The FSARU described the Diablo Canyon seismic design bases and safety analyses results, including assumptions and bounding conditions. This information was used to by the NRC to approve and maintain the facility Operating License.

Description of the safety analysis used to determine the SSE/DDE ground motion.

The safety analysis was compliant with 10 CFR 100, Appendix A.

Included all epicenters within 200 miles and faults within 75 miles of the plant.

The LTSP was completed in 1988.

The LTSP did not address or alter the plant CLB.

The LTSP was not included in the FSARU because the information is not part of the seismic design basis or supporting safety analysis.

The safety analysis considered all active faults passing within 200 miles from the plant when determining the maximum Earthquake for the facility.

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The Diablo Canyon seismic design bases was based on a magnitude 7.25 earthquake on the Nacimiento fault, 20 miles from the site (Earthquake B), and a magnitude 6.75 aftershock associated with a large earthquake on the San Andreas fault (Earthquake D).

The safety analysis did not include consideration of the Hosgri fault when determining the maximum earthquake for the facility. The Hosgri Evaluation (HE) is described as a response to an NRC question, not part of the SSE/DDE design basis.

The safety analysis concluded the maximum peak ground acceleration would be about 0.2 g (grounded at 100 Hz). PG&E designated the SSE/DDE at twice this value, or 0.4 g (grounding at 100 Hz). This approach was accepted by the NRC as equivalent to 10 CFR 100, Appendix A.

Hosgri Evaluation (HE)

The Hosgri fault was discovered a few miles off shore during plant construction by oil company geoscientists. During the Diablo Canyon licensing reviews, PG&E argued that the Hosgri was not a capable, fault as defined in 10 CFR 100, Appendix A, and was not required to be considered for the plant SSE. The NRC argued that the Hosgri fault should be included in the safety analysis for establishing the maximum earthquake for the site.

The resulting compromise is reflected in the CLB. PG&E provided report separate from the FSAR to address the NRCs question concerning the capability of the plant to safely shutdown following a 7.5 magnitude earthquake on the Hosgri fault. 16 This report detailed the methods, assumptions and acceptance criteria to support the conclusion that the plant could safety shutdown following a Hosgri earthquake. The NRC agreed to PG&Es request to use different methodologies, assumptions, and acceptance criteria for the HE. In most cases, these methods and assumptions were less conservative than those approved for the SSE/DDE. The end result was that the Hosgri fault was excluded (exempted) from the GDC 2 SSE design basis.

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The Diablo Canyon FSARU establishes the CLB regulatory and design basis requirements for SSC seismic qualification.

Diablo Canyon complied with 1967 GDC 2 and 10 CFR 100, Appendix A. PG&E also stated that the facility conformed to Part 50, Appendix A, GDC 2 (see Endnote 11 and the Appendix to this DPO).

The DDE is equivalent to the 10 CFR 100, Appendix A, SSE.

PG&E committed to Safety Guide 29, Seismic Design Classification, (Regulatory Guide (RG) 1.29), to determine the set of SSCs required to be seismically qualified for the SSE/DDE. RG 1.29 provided an NRC acceptable method for this determination. The licensee could have proposed a different set of SSCs, subject to NRC approval.

Defines the plant quality, seismic, and design classifications.

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LTSP did not alter or change the Diablo Canyon design bases.

Seismic qualification is based on the (DE/OBE & SSE/DDE) design basis and the HE. In addition to ground motion, the design basis includes the associated analytical methods, initial conditions, etc., applied to each analysis.

Safety analysis results for maximum ground acceleration and response spectra - Earthquakes B or D-modified. This established the seismic design basis controlling parameter as defined in NEI 97-04.

The DE (design earthquake) is equivalent to the operational bases earthquake (OBE) defined in 10 CFR100, Appendix A. The OBE has about 1/2 the peak ground motion of the DDE/SSE.

The safety analysis defined the SSE/DDE as meeting the 10 CFR 100, Appendix A, design basis (the HE was excluded from this analysis).

The FSARU refers to the HE as an answer to an NRC question during the original plant licensing process.

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Discussion of the HE The FSARU refers to the HE as an answer to an NRC question during the original plant licensing process.

The assumptions and methods used for the HE were based on agreements made at meetings with NRC.

The HE demonstrated that the plant could safety shutdown following a 7.5 M earthquake on the Hosgri fault.

The FSARU again clarified that the DDE is the Diablo Canyon SSE and the list of SSCs to be seismically quailed to the SSE are compliant with Guide 1.29, Seismic Design Classification.

In response to the NRC question, the HE established the scope of equipment needed be qualified for safe shutdown following an earthquake on the Hosgri fault. The HE safety functions are different than the specified by Part 100, Appendix A 12

Damping Values Damping values (design basis supporting information) are used in the safety analysis and the HE to calculate how seismic energy attenuates through plant structures and components. Generally, the lower the damping value assumed, the larger amount of seismic stress attenuated through the plant. These damping values are part of the CLB.

NRC approval of the damping values used in the analysis was part of the licensing process. The NRC provided acceptable damping values in Regulatory Guide 1.161, Damping Values for Seismic Design of Nuclear Power Plants. Licensees may use previously NRC approved damping values, for a given material and application, or request approval for alternate values through the license amendment process.

Diablo Canyon Seismic Qualification is Not Limited by the HE Figure 1 illustrates the results of the different methods and assumptions use in DDE/SSE safety analysis and the HE. This figure compares acceleration levels (shaking) in the reactor containment building.

Plant SSCs are most affected in the 3 to 8.5 Hz frequency range.

Note that the level of shaking is significantly greater for the SSE/DDE than for the HE at this plant location.

This may seem counterintuitive since the HE is a much larger earthquake.

However, as this figure illustrates, comparing ground motion alone is not sufficient to evaluate seismic qualification. Methods, assumptions, initial conditions, and acceptance criteria used in the analyses are just as important as ground motion.

Figure 1, Comparison of the DDE/SSE and the HE Floor Response Spectrum, Containment Elevation 88 13

The qualification process used information, such as shown in Figure 1, to establish the amount of seismic stress SSCs may be exposed to during the SSE. A component located at this location would be qualified for the SSE/DDE. If the SSC was also credited for HE safe shutdown, no additional qualification would be required. At this plant location, the seismic stress is dominated by the SSE/DDE. Qualification to the SSE/DDE would envelope the seismic stress generated by the HE.

The FSARU includes many examples where SSC seismic qualification was more limiting by the SSE/DDE than for HE.

In these cases, the SSE/DDE predicts greater seismic stress (shaking) at these plant locations.

Steam generator nozzles Reactor coolant pumps Replacement reactor head 14

American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Requirements are Not Limited by the HE Title 10 of Code of Federal Regulations, Part 50.55a, Codes and Standards, requires important to safety pressure vessels (including the reactor coolant pressure boundary),

system piping, and pipe supports to meet the ASME Boiler and Pressure Vessel Code requirements. Section (iii) of the Rule, Seismic Design of Piping, provides for use of Code Subarticles NB-3200, NB-3600, NC-3600, and ND-3600. These subparts required SSE/DDE seismic loads to be included when verifying plant SSCs meet the Code acceptance criteria. The Code provides assurance that these SSCs important to safety will remain intact following postulated accidents and events, including the SSE/DDE.

The FSARU stated that Diablo Canyon met code requirements (an earlier version of the Code is applicable in some cases)

The CLB requires the Code acceptance limits to be met for SSE/DDE loads combined with accident loads.

HE load combinations and limits were negotiated.

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HE load combinations and some limits were negotiated.

The HE stress limits were relaxed for some Class A components The Code methodology adds seismic loading, generated by either the SSE/DDE safety analysis or the HE, to other non-seismic loads affecting the component. The resulting SSE/DDE stress is significantly greater than for the HE in many loading cases. Again, this may sound counterintuitive since the HE is based on a much larger earthquake. These differences in component stress reflect the differences in the methods, assumptions, load combinations, and initial conditions used in each seismic analysis. For example, Figures 2 and 3 compare the Code bending moments calculated for the control rod drive mechanisms used to support the replacement reactor head modification. As seen in these figures, the bending moments (seismic stress) were much greater for SSE/DDE case than for the HE.

Figure 2 Figure 3 17 18 HE Maximum CRDM Bending Moments SSE/DDE Maximum CRDM Bending Moments 3.0 Concept of Operability The Diablo Canyon Technical Specifications are an attachment to the facility Operating License. 19 The technical specifications include a set of limiting conditions for operation (LCOs) for key plant SSCs. These LCOs are the lowest functional capability or equipment performance level required to ensure safe operation of the facility. When a limiting condition 16

for operation is not met, PG&E is required to shut down the reactor or follow any prescribed remedial actions until the condition can be met. Compliance with technical specification LCOs provide confidence that plant operation is within the boundary of key assumptions used in the safety analysis and preserve the validity of the design bases.

For example, the plant design bases require two redundant trains of emergency core cooling equipment. The safety analysis concluded that either train is capable of successfully mitigating a loss of coolant accident. However, the plant design bases also assume that one train will fail to perform the safety function. Technical Specification LCO 3.5.2 (below) preserves the integrity of these assumptions by ensuring at least one emergency core cooling train will always be available for accident mitigation during plant operation. This LCO limits reactor operation to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when one emergency core cooling train is inoperable and for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> when both trains are inoperable.

To be considered fully qualified, 20 the emergency core cooling system must conform to all aspects of the CLB, including all applicable codes and standards, design criteria, safety analyses assumptions, specifications, and licensing commitments. In contrast, the system is considered degraded or nonconforming when it fails to conform to one or more aspect of the CLB.

An unanalyzed condition exists when the licensee identifies that the plant may be operating outside the bounding conditions assumed in the approved safety analysis.

Power reactor licensees sometimes identify degraded, nonconforming, or unanalyzed conditions that call in to question the capability of plant SSCs to perform the safety functions described in the CLB.

When this occurs, licensees are expected to immediately evaluate the operability of the affected SSCs.

To be considered operable, plant SSC must be capable of performing the safety functions specified by the design, within the required range of design physical conditions, initiation times, and mission times. For operability determination 17

purposes, the mission time is the duration of SSC operation that is credited in the design basis. 21 While this determination may be based on limited information, the information is required to be sufficient to conclude a reasonable expectation that the SSC is operable. If unable to conclude this, the licensee is required to declare the SSC inoperable and apply the technical specification required actions. If the available information is incomplete, the licensee is required to promptly collect any additional information that is material to the determination and promptly make an operability determination based on the complete set of information. If, at any time, information is obtained that negates a previous determination that the SSC is operable, then the licensee is required to immediately declare the SSC inoperable.

For example, a licensee may identify that an incorrect heat transfer coefficient was used in an emergency core cooling performance calculation. This would be considered a nonconforming condition because NRC regulations require that the design basis be correctly translated into supporting design calculations. An operability determination is required because the error calls into question the capability of the system to remove the post-accident heat assumed in the design bases. The licensee would be required to either demonstrate that the specified safety function for the system could still be met, accounting for the effect of the incorrect coefficient, or apply the actions specified in Technical Speciation LCO 3.5.2.

The NRC defines specified safety functions as those safety function(s) described in the CLB for the facility. 22 In addition to providing the specified safety function, a system is expected to perform as designed, tested and maintained. When plant SSC capability is degraded to a point where it cannot perform, with reasonable expectation, or reliability, plant operators are required to consider the SSC inoperable, even if at this instantaneous point in time the system could provide the specified safety function.

The NRC requires the resident inspector to review between 19 and 25 operability evaluations each year at Diablo Canyon. 23 The inspector is asked to verify that degraded or nonconforming SSCs, or compensatory measures taken, does not result in conditions outside of the design basis or inconsistent with safety analyses assumptions.

Summary

a. The plant design bases includes the functions that SSCs are:

(1) required to comply with, including regulations, and license conditions, and (2) credited in the safety analysis to meet NRC requirements.

b. The design base includes the bounding conditions under which SSCs must operate following any accident or event specifically addressed in the CLB.

c. At Diablo Canyon, the SSE/DDE implements the design bases requirements specified in GDC 2 and Part 100, Appendix A. This design basis requires certain SSCs to remain functional following the earthquake which produces the maximum vibratory ground motion for the site, considering the regional and local geology and seismology. These SSCs are those necessary to assure; 18

(1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures (10 CFR 50.34 and 10 CFR 100)

d. SSE/DDE ground motion for the is defined as a design basis controlling parameter.

e. An earthquake on the Hosgri fault was an NRC approved exception to SSE/DDE design basis. While the Hosgri earthquake ground motions exceed those developed for the DDE, PG&E was not required to include the Hosgri fault in the safety analysis for determining the Part 100, Appendix A, maximum earthquake potential for the site.

f. The licensee developed the HE using different methodologies, assumptions, initial conditions, and acceptance criteria, than those approved for the SSE/DDE design basis.

These methods were not included in the FSARU because they were not part of the safety analysis supporting the seismic design basis. Even though the HE represents a larger ground motion, the evaluation is not bounding for Diablo Canyon seismic qualification. In many cases, plant seismic qualification was more limited by the SSE/DDE.

g. The safety analysis demonstrates that SSCs important to safety (listed in RG 1.29 &

Table 1) are capable of performing the specified safety functions and meeting the SSE/DDE design basis. Meeting ASME and other Code acceptance limits provides assurance that pressure retaining systems, including the reactor coolant pressure boundary and containment, will remained intact following a SSE/DDE.

4.0 Chronology Discovery of new Seismic Information November 2008: Pacific Gas and Electric notified the NRC 24 of discovery of a previously unknown zone of seismicity located about a mile offshore from the Diablo Canyon facility.

The licensee stated that an initial assessment indicated that the ground motion from the potential fault was expected to be bounded by the LTSP spectrum. The licensee concluded an operability evaluation was not required because the new information was bound by the LTSP design basis. 25 Initial NRC Review of the Shoreline Fault April 8, 2009: The NRC issued Research Information Letter 09-001, Preliminary Deterministic Analysis of Seismic Hazard at Diablo Canyon NPP from Newly Identified Shoreline Fault to the public. 26 The Research Information Letter included a confirmatory analysis concluding that potential ground motion from the Shoreline fault was bound by the LTSP spectrum. The Research Information Letter did not draw any conclusions related to the Shoreline fault ground motion being within Diablo Canyon CLB. However, the Office of Nuclear Reactor Regulation (NRR) transmittal letter included the following statements:

PG&E informed the NRC staff that it had performed an initial evaluation of the potential ground motion levels at the DCPP from the hypothesized fault which concluded that these motions would be bounded by the ground motion levels previously determined for the current licensing basis.

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Based on the NRC staff review of the preliminary geophysical data provided by PG&E in preparation for the call and the licenses preliminary analysis provided during the conference call, the NRC staff concluded that the current licensing basis is bounding and continues to support safe operation of the DCPP.

Therefore, based on the currently available information, the NRC staff concludes that the design and licensing basis evaluations of the DCPP structures, systems, and components are not expected to be adversely affected and the current licensing basis remains valid and supports continued operability of the DCPP site.

December 15, 2009: Pacific Gas and Electric determined that that the Shoreline Fault was only 300 meters from the plant inlet (location of SSCs important to safety). PG&E again concluded that a nonconforming condition did not exist because the results were still bounded by the LTSP. 27 NRC Discovery of Nonconforming/Unanalyzed Condition September 14, 2010: The resident inspectors identified that postulated Shoreline fault ground motions were greater than those assumed in the DDE safety analysis. 28 The inspectors questioned SSC operability because the DDE was identified as the facility SSE in FSARU Sections 2.5 and 3.7. The inspectors also identified that the LTSP was not part of the seismic design basis.

September 28, 2010: The resident inspectors identified and communicated to PG&E that the Shoreline Fault was a condition outside the bounds of the FSARU seismic safety analysis and was required to be evaluated for operability as defined in station procedures.

PG&E did not take any corrective actions.

October 4, 2010: The resident inspectors recommended an unresolved item be included in Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2010004 and 05000323/2010004, to document concern that an earthquake produced on the Shoreline fault could produce ground motions greater than those described in the SSE/DDE safety analysis. Region IV disapproved the resident inspectors recommendation.

October 5, 2010: The resident inspectors briefed the Office of NRR Project Manager and Branch Chief on the Shoreline fault findings.

October 10, 2010: Pacific Gas and Electric reviewed the inspectors operability concerns prior to releasing Unit 1 for restart following refueling. Pacific Gas and Electric again concluded that a nonconforming condition did not exist because predicted ground motions were within the LTSP spectrum. 29 October 14, 2010: The resident inspectors briefed the Region IV Regional Administrator on the Shoreline Fault findings.

Pacific Gas and Electrics Failure to Assess Operability October 19, 2010: The resident inspectors met with the PG&E engineering vice president and discussed seismic operability concerns. The engineering vice president stated that the problem was related to an incomplete plant licensing docket. The vice president argued 20

that past agreements made with the NRC to only use the LTSP to evaluate new seismic information were inadvertently omitted from docketed correspondence and the FSARU. The vice president also stated that no additional action was required because the Shoreline fault spectrum was bound by the LTSP.

November 30, 2010: The resident inspectors provided a detailed briefing of the Shoreline fault findings to the Region IV, Reactor Projects Division Director. At this meeting, the Reactor Projects Deputy Division Director took the action to request the PG&E engineering vice president to enter the Shoreline fault into the corrective action program and assess the effect of the higher ground motions on plant SSC (perform an operability evaluation).

December 16, 2010: Pacific Gas and Electric again declined to evaluate operability of plant SSCs. PG&E engineering and regulatory assurance staff indicated that the Shoreline fault ground motions were too high to successfully demonstrate SSCs operability using the SSE/DDE methods specified in the CLB. In response to the Deputy Division Directors request, PG&E updated the condition report to include a justification for not evaluating the operability of technical specification required SSCs. 30 This justification included a summary of the April 8, 2009 NRC NRR letter:

Therefore, based on the currently available information, the NRC staff concludes that the design and licensing basis evaluations of the DCPP structures, systems, and components are not expected to be adversely affected and the current licensing basis remains valid and supports continued operability of the DCPP site.

January 2011: PG&E submitted a report to the NRC updating the local geology. 31 This report included detailed deterministic evaluations of the San Luis Bay, Los Osos and Shoreline faults. The report concluded that each of these faults are capable of producing significantly greater vibratory ground motion than assumed in the SSE/DDE safety analysis (Table 2). The inspectors concluded that this information resulted in an unanalyzed condition because the new predicted ground motions where greater that those used as bounds for the existing SSE/DDE safety analysis and seismic qualification basis. The inspector again recommended that Region IV initiate enforcement action because PG&E had failed to demonstrate that technical specification required SSCs were capable of performing the required safety functions. 32 The inspector included a second enforcement recommendation to address the incomplete and inaccurate information PG&E provided the NRC related to the seismic design basis. This incomplete and inaccurate information lead to the incorrect conclusions stated in the April 8, 2009 NRC NRR letter.

Table 2 Comparison of Reanalysis to Diablo Canyon SSE Local Earthquake Fault 33 Peak Ground Acceleration 34 SSE/DDE Design Basis 0.40 g Shoreline Faults 0.62 g Los Osos 0.60 g San Luis Bay 0.70 g Hosgri (HE) 0.75 g Note: Peak ground acceleration is anchored at 100 Hz and only used as a bases for comparison 21

NRC Initial Response to Seismic Operability April 2011: The resident inspector met with the NRR Project Manager, NRR Branch Chief and the Region IV, Reactor Projects Division Director. The inspector again recommended that the NRC initiate enforcement action against PG&E. Enforcement action was required because the licensee continued to operate the plant outside the bounds of the safety analysis. The licensee had refused to demonstrate SSC operability at the higher ground motions or shutdown the reactors in accordance with technical specifications. At the meeting, Reactor Projects Division Director stated that initiating enforcement action would reverse the previous NRC conclusion described in the April 8, 2009 NRR letter, that the new seismic information was within the facility design basis. The Division Director requested that NRR formally concur on this reversal of position prior to the agency initiating action. At the Division Directors request, the inspector initiated a Task Interface Agreement to document NRR concurrence on the new position.

May 2011: The NRC opened Unresolved Item: 05000275; 323/2011002-03, Requirement to Perform an Operability Evaluation Following Receipt of New Seismic Information. 35 This Unresolved Item identified NRC concerns that PG&E had failed to evaluate the effect the new seismic information had on capability of plant SSC to perform the requires safety functions at the higher seismic stress:

The inspectors were unable to confirm the licensees statements that new seismic information was only required to be evaluated under the LTSP deterministic margin analysis (which is a margin analysis to the Hosgri Event) based on a review of docketed information and the plant safety analysis. The LTSP margin analysis only demonstrated that the new seismic information was bound by the Hosgri Event design basis earthquake, not the Design or Double Design Earthquakes.

August 2011: The NRC issued Task Interface Agreement (TIA) 2011-010, Concurrence on Diablo Canyon Seismic Qualification Current Licensing and Design Basis. 36 This TIA documented the agency position that new seismic information developed by the licensee was required to be evaluated against the design earthquake (DE), the DDE, and HE, including the assumptions used in the supporting safety analyses as described in the FSARU. The staff concluded that comparison only against the LTSP (a margin analysis to the HE) was not sufficient to meet this requirement.

October 2011:

• Pacific Gas and Electric completed an operability evaluation of the effect of the new seismic information. The licensee concluded that all plant technical specification SSCs were operable because the new ground motions were less than those assumed in the HE. The licensee stated that based on engineering judgment, the HE was sufficient to satisfy SSE/DDE design basis requirements for operability.

• Pacific Gas and Electric requested NRC approval to change the Diablo Canyon SSE design basis from the DDE to the HE (License Amendment Request 11-05). 37 The licensee submitted the amendment request following several NRC meetings at which various approaches for incorporating the new seismic information into the CLB were discussed.

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December 2011: Pacific Gas and Electric submitted Letter DCL-1 1-124, Standard Review Plan Comparison Tables for License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake, to the NRC. 38 This letter included 66 attachments (320 pages) detailing the deviations and exceptions between the HE methodology and the NRC SSE review standards (NUREG 800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition). The NRC had requested this information to aid in the acceptance review of License Amendment Request 11-05.

January 2012: The resident inspector concluded that the PG&E October 2011 operability determination failed to meet NRC inspection standards. The inspector based this conclusion on:

• The operability determination failed to demonstrate that all ASME Code requirements were met for the higher ground motions. The licensees failure to demonstrate Code compliance called in to question the integrity of the reactor coolant pressure boundary following an earthquake on the Los Osos, San Luis Bay or Shoreline faults.

• The operability determination failed to demonstrate that all plant SSCs credited in the in the SSE design basis would remain functional at the higher stress levels represented by the new ground motions. The licensees comparison of the new ground motions only against the HE was not adequate to demonstrate that SSE/DDE CLB requirements were satisfied.

The inspector again recommended that the agency initiate enforcement action against PG&E based on the licensees failure to demonstrate that technical specification required equipment would remain function at the higher ground motions. The agency disagreed with the inspectors recommendations (documented in non-concurrence NCP-2012). 39 The staff stated that the licenses comparison of the new seismic information against the HE was adequate to demonstrate initial operability. The staff also stated that additional review of Licensee Amendment Request 11-05 was needed before the agency had enough information to complete an operability determination.

February 2012:

• The NRC issued non-cited violation, 05000275; 323/2011005-02, Failure to Perform an Operability Determination for New Seismic Information. 40 This violation addressed the failure of PG&E to initially perform an operability determination following development of the new seismic information back in January 2011.

• The NRC closed Unresolved Item: 05000275; 323/2011002-03. 41 The staff concluded that PG&E corrective actions were adequate to conclude all Diablo Canyon SSCs were operable:

The staff concluded that 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. In order to complete a comprehensive evaluation, the licensee needed NRC approval of the methodology to be used to complete this evaluation.

September 2012: The resident inspector was reassigned from Diablo Canyon 23

Subsequent NRC Actions to Address New Seismic Information October 2012:

• The NRC completed an evaluation of the Shoreline fault. The staff concluded that the Shoreline scenario should be considered as a lesser included case under the HE. 42 The NRC stated:

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. 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).

• At the NRCs request, PG&E withdrew License Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake." 43 The license amendment request had not met the NRCs acceptance review standard.

November 2012: The NRC revised Task Interface Agreement (TIA 2011-010) Diablo Canyon Seismic Qualification Current Licensing and Design Basis. 44 The revised TIA stated:

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 Update, as necessary, to include the Shoreline scenario in accordance with the requirements of 10 CFR 50.71(e).

The NRCs letter dated October 12, 2012, and the request for information dated March 12, 2012, (50.54(f)) provide guidance for assessing new seismic information and what PG&E is expected to do in the event that it becomes apparent that the new seismic information will lead to a GMRS that is higher than the DDE.

5.0 NRC Corrective Actions to Address Deficient Seismic Safety Analysis were Inadequate The Staff Proposed FSARU Update Requires an Amendment to the Diablo Canyon Operating License The staff recommended that PG&E update the FSARU to include the Shoreline scenario as a lesser included case of the HE. 45 This change exempts the Shoreline fault from the existing SSE/DDE design basis requirements. PG&E is required to review proposed FSARU updates under the provisions of 10 CFR 50.59, Changes, Tests and Experiments. 46,47 This review determines if the proposed change will require an NRC approved amendment to the Operating License prior to implementation. 10 CFR 50.59 states a license amendment is required for changes that:

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the FSARU, or 24

Result in a departure from a method of evaluation described in the FSARU used in establishing the design bases or in the safety analyses Title 10, Code of the Federal Regulations, Part 50.59, includes the following definitions:

• Change: A modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

• Departure from a method of evaluation described in the FSARU used in establishing the design bases or in the safety analyses:

Changing any of the elements of the method described in the FSARU unless the results of the analysis are conservative or essentially the same; or Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.

• Facility as described in the FSARU:

The structures, systems, and components that are described in the FSARU, The design and performance requirements for such SSCs described in the FSARU, and The evaluations or methods of evaluation included in the FSARU for such SSCs which demonstrate that their intended function(s) will be accomplished.

• Tests or experiments not described in the FSARU means any activity where any SSC is utilized or controlled in a manner which is either:

Outside the reference bounds of the design bases as described in the FSARU or Inconsistent with the analyses or descriptions in the FSARU.

The 50.59 requirements are expanded in the NRC endorsed guidance contained in Nuclear Energy Institute, NEI 96-07, Guidelines for 10 CFR 50.59 Evaluations, Revision 1: 48,49 Adding the Shoreline scenario to the FSARU HE analysis would result in more than a minimal increase in the likelihood of a malfunction of plant SSC because the change departs from the design basis requirements established by GDC-2. NEI 96-07 states:

Section 4.3.2 Does the Activity Result in More than a Minimal Increase in the Likelihood of Occurrence of a Malfunction of an SSC Important to Safety?

The term "malfunction of an SSC important to safety" refers to the failure of SSC to perform their intended design functions-including both non-safety-related and safety-related SSCs. The cause and mode of a malfunction should be considered in determining whether there is a change in the likelihood of a malfunction.

In determining whether there is more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC to perform its design function as described in the UFSAR, the first step is to determine what SSCs are affected by the proposed activity. Next, the effects of the proposed activity on the affected SSCs should be determined. This evaluation should include both direct and indirect effects.

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Changes in design requirements for earthquakes, tornadoes, and other natural phenomena should be treated as potentially affecting the likelihood of malfunction.

Although this criterion allows minimal increases, licensees must still meet applicable regulatory requirements and other acceptance criteria to which they are committed (such as contained in Regulatory Guides and nationally recognized industry consensus standards, e.g., the ASME B&PV Code and IEEE standards). Further, departures from the design, fabrication, construction, testing, and performance standards as outlined in the General Design Criteria (Appendix A to Part 50) are not compatible with a "no more than minimal increase" standard.

The Shoreline Scenario results in SSC seismic stress beyond the plant SSE qualification basis. Exposure to higher levels of stress results in an increases likelihood of a malfunction of these SSCs. The change also increases the likelihood of a malfunction of SSCs important to safety because removing the Shoreline scenario from the SSE/DDE departs from applicable regulatory requirements and other acceptance criteria the PG&E had committed to for the SSE/DDE.

The staff proposed FSARU update also requires a licensee amendment because applying the HE methodology to Shoreline fault changes the methods described in the FSARU for establishing the SSE design basis. NEI 96-07 states:

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?

The UFSAR contains design and licensing basis information for a nuclear power facility, including description on how regulatory requirements for design are met and how the facility responds to various design basis accidents and events. Analytical methods are a fundamental part of demonstrating how the design meets regulatory requirements and why the facility's response to accidents and events is acceptable. As such, in cases where the analytical methodology was considered to be an important part of the conclusion that the facility met the required design bases, these analytical methods were described in the UFSAR and received varying levels of NRC review and approval during licensing.

As discussed further below, for purposes of evaluations under this criterion, the following changes are considered a departure from a method of evaluation described in the UFSAR:

• Changes to any element of analysis methodology that yield results that are non-conservative or not essentially the same as the results from the analyses of record.

• Use of new or different methods of evaluation that are not approved by NRC for the intended application.

As described in the FSAR Section 2.5, the seismic SSE/DDE design basis includes the shoreline scenario because the fault is located within 75 miles of plant site. The HE was an exception to this design basis. To change the plant safety analyses to also exclude the Shoreline scenario from the seismic design basis results in a departure from a method described in the FSARU that was used to establish the SSE/DDE design basis. NRC approval, in the form a license amendment, is required before the HE methods, including assumptions, initial conditions, etc., can be applied to other local seismic features.

The licensee previously requested that the NRC approve the new information as part of the HE (License Amendment Request 11-05). 50 However, the NRC did not accept the license 26

amendment request for review. The NRC standard for acceptance review required that the license amendment request demonstrate that the proposed change would not impose a significant hazard.

The NRC corrective action was also inadequate because the disposition of the San Luis Bay and Los Osos faults was omitted. PG&E had determined that these faults also had significant impact on plant equipment. The FSARU SSE safety analysis is also nonconforming with respect to the deterministic evaluations of the San Luis Bay and Los Osos faults.

Existing Regulatory Framework Title 10 of the Code of Federal Regulations, Parts 50.34 and 50.71(e), required PG&E to include information in the FSARU that describes the facility, presents the design bases and the limits on its operation, and present a safety analysis of the SSCs and of the facility as a whole. These regulations define safety analyses as analyses performed pursuant to NRC requirement to demonstrate:

(1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in 10 CFR 50.34(a)(1).

The safety analysis is required to demonstrate that acceptance criteria for the facility's capability to withstand or respond to postulated events are met. Supporting FSARU analyses are required to demonstrate that SSC design functions will be accomplished as credited in the accident analyses of events that the facility is required to withstand such as earthquakes and accidents. As previously discussed, the new seismic information resulted in the existing FSARU safety analysis nonconforming with the design basis and Parts 50.34 and 50.71.

Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control required PG&E to maintain the plant configuration consistent with regulatory requirements and the design basis:

Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in § 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. These measures shall include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from such standards are controlled.

A violation of Criterion III occurred after PG&E concluded that the new seismic information would produce greater ground motion that bound by the plant SSE safety analysis and design bases (established by GDC 2 and Part 100). Design measures no longer provided assurance that the important to safety SSCs are capable of performing the required safety functions at the higher ground motions.

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10 CFR 50, Appendix B, Criterion XVI, Corrective Action, required PG&E to implement prompt corrective action to restore the plant as described in the safety analysis and design basis:

Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to appropriate levels of management.

A violation of Criterion XVI occurred after PG&E failed to take prompt corrective actions to correct deficiencies in the plant safety analysis, as required by 10 CFR 50.34 and 50.71(e) and to restore plant SSCs within the capability of meeting the seismic design basis as required by Appendix B, Criterion III .

No Viable Corrective Action Path This regulatory framework ensures that licensees promptly restore plant operation within the boundary of the design basis and NRC approved safety analysis. Changing the local seismology to meet the CLB is beyond the licensees control. Adapting plant SSCs to meet the current design basis requirements, if even possible, would require extensive seismic retrofits. Modifying the design basis and safety analysis to accommodate the new information would require an amendment to the Operating License. However, the NRC was not willing to accept the amendment request for review. The end result is the licensee is without a viable corrective action path to deal with the current nonconforming and unanalyzed conditions. The lack of a clear corrective path does not waive the NRCs responsibility to enforce current regulatory requirements for prompt corrective actions and to ensure plant operation is maintained within the boundaries of the approved safety analysis.

Fukishima Near-Term Task Force 10 CFR 50.54(f) Requested Information is not Applicable to the Current Diablo Canyon Nonconforming and Unanalyzed Conditions In March 2012, the NRC requested information related to the reevaluation of seismic hazards at all power reactor facilities. 51 This request was in response to recommendations from the NRC Near-Term Task Force review of the Fukishima accident. The NRC requested that PG&E develop new probabilistic ground motion models and compare the results of these models to the existing deterministic SSE/DDE. This comparison will provide risk information related to the local geology. The agency will use this risk based information to make future licensing decisions.

The requested information is probabilistic in nature. The Diablo Canyon design bases are deterministic in nature, assuming that the event occurs and requirement specific acceptance criteria are met. While the requested 50.54(f) information will provide risk insights to earthquake hazards affecting the plant, this information is not directly relevant to the CLB.

In contrast, the new deterministic information developed by PG&E for the San Luis Bay, Los Osos, and Shoreline faults was directly comparable to the existing facility design bases and Operating License. This new information was sufficient to conclude that the plant is operating outside of the NRC approved safety analysis and the design bases. The current regulatory framework requires these nonconforming and unanalyzed conditions to be 28

promptly disposition within the context of the CLB. These actions are required independent of information developed in response to the 50.54(f) request.

Summary Pacific Gas and Electric submitted to the NRC information concluding that three local earthquake faults are capable of producing greater ground motion than bounded by the NRC approved safety analysis and the design basis. This condition rendered the plant seismic safety analysis nonconforming with NRC regulations. The NRC has failed to enforce quality requirements (Part 50, Appendix B) that required the licensee to take prompt action to correct the nonconforming safety analysis.

The Staff recommended that PG&E updated the FSARU to include one of these faults as a lesser case under the HE. This action bypassed the regulatory processes (50.2 & 50.90) design to ensure that these changes would not result in a significant hazard. NRC regulations (50.59) require that the licensee first obtain a license amendment before updating the FSARU with this information. A license amendment is required because this change attaches the same regulatory dispensation approved for the Hosgri to the Shoreline fault. The staffs conclusion that reasonable assurance of safety is not an adequate basis to bypass the regulatory requirements to amend the facility Operating License.

The licensee previously submitted a license amendment request to redefine the HE as the SSE for the facility. However, this request did not meet the NRCs minimum standards for acceptance into the review process. As a result, the Staff requested that PG&E withdraw the request.

Deferral of corrective action pending completion of the Fukishima Near-Term Task Force seismic reviews is inconsistent with the current regulatory framework. The new seismic information generated by the licensee was sufficient to conclude that the facility is currently operating outside of the current safety analysis and design basis.

The staffs corrective action was also deficient because the reevaluation of the San Luis Bay and Los Osos faults was omitted. While these faults were initially evaluated in the LTSP, the licensee had not deposition the effect of the higher ground motions on the SSE/DDE safety analysis as required by NRC quality regulations. The SSE/DDE safety analysis is also nonconforming due to the higher ground motions associates with these faults.

6.0 The NRC has not Verified Plant Technical Specification Required SSCs are Operable Plant operators are required to demonstrate that all affected technical specification required SSCs are operable following identification of nonconforming or unanalyzed conditions.

The operability processes provide a basis that the reactors can be operated safely during the corrective action period.

Applicability of Operability Process A nonconforming condition exists because the Diablo Canyon FSARU safety analysis is no longer compliant with the regulatory requirements of GCD-2 for earthquakes. NRC operability policy states: 52 29

Failure to meet a GDC in the CLB should be treated as a degraded or nonconforming condition and, therefore, the technical guidance in this document is applicable.

Also, this was an unanalyzed condition because the new information indicated that the ground motions assumed in the SSE/DDE safety analysis (earthquakes B & D) were no longer bounding for the plant seismic qualification basis. Nonconforming or unanalyzed conditions that call into question the capability of technical specification required SSCs to perform the specified safety functions are required to be evaluated for operability. 53 Description of NRC Operability Process The applicable CLB requirements for seismic qualification must be identified before operability can be evaluated. The new deterministic ground motions were applicable to the SSE/DDE safety analysis, as described in FSARU Section 2.5 and 3.7, because:

• The new seismic information was identified on earthquake faults within 75 miles from the plant.

• The new seismic information was not associated with the Hosgri fault (the NRC approved exception).

• The SSE/DDE safety analysis implemented the plant seismic design basis, and License and regulatory requirements.

Engineering Margins The operability process allows licensees to use engineering margins. Engineering margins include the difference between actual SSC capability and the performance requirements specified in the CLB. To illustrate this concept, consider the emergency core cooling system example discussed in Sections 2.0 and 3.0. This system has motor operated valves and instruments located around the 88 foot elevation level in the containment building. The seismic stress used to develop the original qualification of these SSCs was shown in Figure 1. The new seismic information calls into question the operability of these SSCs because an earthquake on the San Luis Bay fault would result in much higher vibratory motions at this plant location than considered in the SSE/DDE safety analysis. The design basis remains unchanged; these SSCs still are required to remain functional following the A comparison of the new seismic information against the existing SSE/DDE safety analysis would yield seismic stress greater than the values used during the original SSC qualification. However, in many cases, the actual SSC qualification tests were performed at higher levels than required to meet the design basis. These higher qualification levels provide engineering margin that may be recovered for operability.

The operability process does not require that the new ground motions be reviewed against the HE (red line). As described in the CLB, the HE is limited to an earthquake on the Hosgri fault. Also, at this plant location, seismic qualification would Figure 4 likely be bound by the DDE rather than HE.

Comparison of the DDE/SSE and the HE Floor Response Spectrum, Containment Elevation 88 30

maximum earthquake. The vibratory motions associated with the maximum earthquake have changed.

Plant components were generally qualified at higher stress levels (shaking) than the limits specified in the design and engineering specifications. The difference between the reevaluated stress and the actual stress levels used to qualify theses SSCs provides engineering margin. Figure 4 compares the postulated increase in vibratory motions from the San Luis Bay fault against the original DDE qualification levels. The SSCs could be considered operable, if the original qualification was bound at the new stress levels.

Operability also provides for the use of alternate methods. The license may present an alternate method that demonstrates that the SSC will remain functional beyond the qualified level of shaking. The NRC standard is a reasonable assurance that the SSC will be capable of performing the required safety functions, as described in the CLB, at the higher vibratory motions. For example, the licensee could provide alternate testing data that demonstrates the SSC would remain functional at the higher vibratory motions.

Use of Code Margins Engineering margin in the ASME Code calculations may be similarly credited for operability. For example, again consider the emergency core cooling system example. To be considered operable, the Code acceptance limits must be met at the higher stress levels for the system piping and pipe hangers. Plant operators may credit the margin between the actual pipe stress and Code acceptance limits. For example, the original DDE calculation may have determined that an emergency core cooling pipe weld had bending moment of 120,000 lbf-in with a Code acceptance limit of 200,000 lbf-in. The original calculation provided 80,000 lbf-in of margin. This margin may be used for operability when the bending moment is recalculated at the higher seismic stress. The component would be considered operable provided the new bending moment is still less than the Code acceptance limits.

Use of Safety Analysis Margins Methods and supporting design information, used in the safety analysis also provide margins that may be recovered in the operability process. For example, consider the affect damping values have on seismic qualification. Energy dissipation within a structure during an earthquake depends on a number of factors, including the types of joints or connections used within the structure, the structural material, and the magnitude of deformations experienced. In a dynamic elastic analysis, this energy dissipation is usually accounted for by specifying an amount of viscous damping. The damping value affects the energy dissipation in the analytical model. Figure 5 shows the relationship between acceleration and velocity as a function of damping. 54 This relationship determines the level of SSC vibratory motion for seismic qualification. Figure 6 illustrates the relationship between the damping value and the predicted attenuation of seismic energy. Generally, the higher the assumed damping value, for a given spectra, the lower the resulting vibratory motion transmitted to the SSC.

FSARU Section 3.7.1.3, Critical Damping Values, specified the damping values used in the SSE/DDE safety analysis. NRC approval of the SSE/DDE safety analysis included comparing these damping values against NRC review criteria. However, these damping values may contain margin that could be recovered in the operability process. The NRC 31

operability policy allows use of engineering judgment. Use of higher damping values would reduce the amount of seismic stress assumed to attenuate to the plant SSCs. Use of engineering judgment is subject to a couple of tests. 55,56 In such instances, the application of the alternative analysis must be consistent with the technical specifications, license condition, or regulation If the analytic method in question is described in the CLB, the licensee should evaluate the situation-specific application of this method, including the differences between the CLB-described analyses and the proposed application in support of the operability determination process.

Occasionally, a regulation or license condition may specify the name of the analytic method for a particular application. In such instances, the application of the alternative analysis must be consistent with the technical specifications, license condition, or regulation.

Figure 5 Figure 6 Relationship between Acceleration, Relationship between Damping &

57 Velocity as a Function of Damping Propagation of Seismic Energy Higher damping values may be used for operability, provided that these values are appropriate to the application, as defined in the CLB. For example, the damping values specified for the SSE in Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, 58 may be used. Also, damping values higher than presented in Regulatory Guide 1.61, may also be used provided that they have been NRC approved for the specific application and material.

Engineering Margins were Insufficient to Demonstrate Operability These NRC principles were not practical for determined SSC operability for the new seismic information. The new vibratory motions are much greater than those bound by the existing SSE/DDE CLB. This combined with very little engineering margin available in the original SSE/DDE safety analysis would likely result in the CLB acceptance criteria to be exceeded.

32

NRC Conclusion all Diablo Canyon Seismically Qualified Equipment were Operable The NRC concluded that all Diablo Canyon technical specification required SSCs were operable after performing a review of new earthquake potential. 59 The staff stated that NRC operability requirements were satisfied because the new ground motions were bound by those assumed in the HE and LTSP. During this review, the staff also stated:

• The NRC will not ask the licensee to use the new ground motion input data in the DE or DDE 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 ground motion data and the calculation method, including damping values, are correlated parameters. They must be based on the same assumptions for the calculation to have validity.

• It is appropriate for the licensee to use the available new ground motion data in the HE analysis because the new ground motion data is consistent with that evaluation.

Operability was not Evaluated Against the Current Design and Licensing Bases The NRC failed to assess operability against the CLB. The staffs approach to exclude the SSE/DDE design basis and safety analysis for the seismic operability determination was not support by NRC operability policy. Operability required that SSC performance be compared against CLB requirements. 60 In order to be considered operable, an SSC must be capable of performing the safety functions specified by its design, within the required range of design physical conditions The CLB includes the SSE/DDE safety analysis. This safety analysis implements the plant seismic design basis and demonstrates specific regulatory requirements are met. The staffs argument for not using the SSE/DDE for operability was that the new seismic loads were beyond the capability and limitations of the safety analysis. In other words, the NRC acceptance criteria cannot be demonstrated when the new ground motions are compared against the plant SSE design basis. When the operability determination fails to demonstrate these specified safety functions can be met, then the system should be considered inoperable. 61 The specified function(s) of the system, subsystem, train, component or device (hereafter referred to as system) is that specified safety function(s) in the CLB for the facility.When system capability is degraded to a point where it cannot perform with reasonable expectation or reliability, the system should be judged inoperable.

The staffs argument is correct that the HE, including assumptions, initial conditions, and acceptance criteria, is more consistent with the new ground motions. The HE methodology may be adapted by the staff as a basis for a licensing action. However, the HE may not be used as a standard for operability because the methodology was not approved for the SSE as described in the CLB. As such, the HE cannot be the basis to conclude SSCs are operable for the SSE design basis.

While the HE damping values and other inputs are correlated parameters, the CLB restricts the use of these values to analysis of the Hosgri fault (FSARU Section 3.7). The CLB prescribes the damping values and other inputs to be used for the SSE. Substitution of HE damping and other inputs for operability, based solely on the magnitude of the new ground 33

motions, is inappropriate. Use of higher damping values is permitted provided the NRC has approved those values for same application (for the SSE and specified materials). The NRC Operability process requires these input values be consistent with those used in the SSE CLB.

As described in Section 4.0, Chronology, the licensee had requested NRC approval to use the HE methodology for SSE applications (License Amendment Request 11-05). 62 PG&E Letter DCL-1 1-124, described the considerable departure between the HE methodology and the NRCs SSE approval standards. 63 The end result was that the NRC did not accept the licensees request for review. The licensee was unable to demonstrate that use of the HE for SSE applications met the no significant hazards consideration standard. 64,65 While not appropriate for operability, use of the HE analysis, and correlated input parameters, may use as a basis for NRC approval of an amendment to the facility Operating License or waving regulatory (50.2, 50.55a) or technical specification requirements.

The NRC Operability Method Over-Predicted SSC Performance when Compared to the CLB NRC policy allows use of alternative analytical methods when performing operability determinations. However, these methods are required to be consistent with the methods used in the CLB and not over-predict the capability of plant SSC. 66 If the analytic method is not currently described in the CLB, the models employed must be capable of properly characterizing the SSCs performance. This includes modeling of the effect of the degraded or nonconforming condition.

Acceptable alternative methods such as the use of best estimate codes, methods, and techniques. In these cases, the evaluation should ensure that the SSCs performance is not over-predicted by performing a benchmark comparison of the non-CLB analysis methods to the applicable CLB analysis methods.

Comparing the new information solely against the HE attaches all of the HE methods and assumptions to the new information. These methods and assumptions result in significantly underestimating the resulting seismic stress that plant SSCs would be exposed when compared to the SSE/DDE methods described in the CLB. As a result, use of the HE over-predicts SSC seismic performance when compared to the SSE/DDE CLB methods.

As discussed in Section 3.0, the SSE/DDE safety analysis predicated greater stress (shaking) and was more limiting for the seismic qualification of some plant SSCs than for the HE. As demonstrated in these examples, ground motion taken alone is not a meaningful representation of the seismic design bases. 67 Considered the control rod drive mechanism bending moment example discussed in Section 3.0, Diablo Canyon Current Licensing Bases. Appling the HE methods to the San Luis Bay ground motions would result in less stress than shown in Figure 2. This is because the San Luis Bay fault spectrum is slightly lower than the HE. However, applying SSE/DDE methods to San Luis Bay fault would result in significantly larger stresses than shown in Figure 3. HE methods are not appropriate for operability because these method significantly over-predict the capability of plant SSCs when compared to the CLB method (SSE/DDE).

34

NRC Operability Review Failed to Demonstrate ASME Code Requirements were Met Title 10, Code of Federal Regulations, Part 50.55a, Codes and Standards, requires the licensee to meet the ASME Boiler and Pressure Vessel Code requirements. The Code requires the SSE maximum earthquake dynamic loading to be included when demonstrating the acceptance limits are met for Class1 systems. The new information concluded that higher vibratory motions could affect plant Code components that were used in the original SSE/DDE calculations. The HE cannot be used for SSE Code compliance because the HE (along with the methods, assumptions, etc.) was not identified as the SSE in the CLB. This new loading calls into question if Code limits can still be met given the potential for a much larger maximum earthquake. Operability requires certain plant SSCs either meet the ASME Code acceptance criteria or provisions in an NRC approved Code Case. 68 When ASME Class 1 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.

Structures may be required to be operable by the Technical Specifications, or they may be related support functions for SSCs in the Technical Specifications..As long as the identified degradation does not result in exceeding acceptance limits specified in applicable design codes and standards referenced in the design basis documents, the affected structure is either operable or functional.

When a degradation or nonconformance associated with piping or pipe supports is discovered, the licensee should use the criteria in Appendix F of Section III of the ASME Boiler and Pressure Vessel Code for operability determinations. The licensee should continue to use these criteria until CLB criteria can be satisfied (normally the next refueling outage). For SSCs that do not meet the above criteria but are otherwise determined to be operable, licensees should treat the SSCs as if inoperable until NRC approval is obtained to use any additional criteria or evaluation methods to determine operability. Where a piping support is determined to be inoperable, the licensee should determine the operability of the associated piping system.

The NRC Inappropriately Deferred Operability Pending License Amendment Request Approval The NRC stated: 69 The staff concluded that 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. In order to complete a comprehensive evaluation, the licensee needed NRC approval of the methodology to be used to complete this evaluation.

NRC operability does not provide for an indeterminate state. 70 Plant SSCs are either operable or inoperable. The operability process also does not include initial basis for Operability. NRC policy only provides for immediate and prompt operability determinations. Prompt operability determinations should be completed within the technical specification out-of-service times. 71 For the seismic issues, this would be about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Operability is assessed against the CLB, not against a pending license amendment request. Plant SSCs should be immediately considered inoperable when 35

inadequate margin is available, as described in the CLB, to ensure the components are capable of performing the CLB specified safety functions. The staffs deferral of comprehensive evaluation for operability was inconsistent with the current regulatory framework and Diablo Canyon Operating License.

Research Information Letter 12-01, "Confirmatory Analysis of Seismic Hazard at the Diablo Canyon Power Plant from the Shoreline Fault Zone" In October 2012, the NRC released Research Information Letter 12-01. 72,73 This Letter included the results of a conformational analysis of potential ground motions that could be produced by the Shoreline fault. The Letter did not address the seismic qualification of plant SSCs, ASME Code requirements, or operability. However, the Letter stated:

It should be reiterated that the NRC staff has concluded 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 HE ground motion and the 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 existing design basis for the plant already is sufficient to withstand those ground motions.

• The staffs conclusion of reasonable assurance of safety is not applicable to either resolving the noncompliant safety analysis or determining operability. This information may be useful input for regulatory decisions, such as approval of license amendments or exemptions from existing regulations. However, the current regularly framework and facility Operating License requirements are still required to be satisfied. Continued operation of Diablo Canyon is dependent on successful demonstration of SSC operability. Since operability is evaluated against the CLB, this demonstration may require amendment of the Operating License and/or waving current regulatory requirements. The staffs conclusion of reasonable assurance of safety may be used to support justification for these regulatory actions.

• The current regulatory framework does not provide for deferral of the operability evaluation until development of new probabilistic ground motions models, such as those requested by the Fukishima Near-Term Task Force. Sufficient information is currently available to assess operability. Because the facility design bases is deterministic in nature, the NRC operability policy specifically excludes use of probabilistic information:74 Probabilistic risk assessment is a valuable tool for evaluating accident scenarios because it can consider the probabilities of occurrence of accidents or external events. Nevertheless, the definition of operability is that the SSC must be capable of performing its specified safety function or functions, which inherently assumes that the event occurs and that the safety function or functions can be performed. Therefore, the use of PRA or probabilities of occurrence of accidents or external events is not consistent with the assumption that the event occurs, and is not acceptable for making operability decisions.

Summary The staff failed to enforce plant technical specification requirements to shut down the Diablo Canyon reactors. Continued reactor operation was dependent on the licensees demonstration that technical specification required SSCs were operability following discovery of nonconforming and unanalyzed conditions associated with the new seismic 36

information. The failure to demonstrate operability, required the licensee to take the prescribed technical specification actions for the inoperable equipment, including shutdown the reactors. The operability determination method used by PG&E was inadequate because:

• Neither the HE nor the LTSP methods were approved by the NRC to be used for the Diablo Canyon SSE design basis. The CLB defined the HE as an exception to the SSE and was only approved for evaluating the Hosgri fault. The LTSP is not part of the seismic design basis.

• Use of the HE and LTSP over-predicts SSC performance when compared to the CLB methods used for the SSE/DDE. Neither the HE nor the LTSP are bounding for SSC seismic qualification at Diablo Canyon. Comparisons limited to only ground motion are meaningless for operability. These comparisons omit other relative CLB requirements including the methods, assumptions, initial conditions, and acceptance criteria applicable to each evaluation.

• Comparison of the new information only to the HE and LTSP failed to demonstrate that the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code are met at the higher ground motions. Operability requires that the Code acceptance criteria are met for key plant components, including the reactor coolant pressure boundary.

The staffs conclusion in Research Information Letter 12-01 that reasonable assurance of safety exists does not provide an adequate basis for concluding operability. A reasonable assurance of safety does not satisfy the requirement that plant SSCs are capable of meeting the specific safety functions described in the SSE/DDE safety analysis and design basis.

7.0 Previous Attempts for Resolution

a. The author of the DPO discussed these issues with senior Region IV management, including the region administrator, and NRR Division of Operating Reactor Licensing staff between the fall 2010 and the fall of 2012 (see Section 4.0, Chronology).

b. The author of the DPO was not provided an opportunity to review or supply input to either the October 2012 NRC letter75 or the revised TIA 11-05. 76

c. The author of the DPO provided written recommendations for regulatory action in January 2011. 77

d. The author of the DPO discussed the definition of deign basis and applicability of 10 CFR 50.59 to the NRC recommend FSARU changes with the Region IV, Division of Reactor Projects, Chief of Reactor Projects Branch B, on June 27, 2013.

e. The author of the DPO non-concurred on Diablo Canyon Power Plant Inspection report 050000275/323-2011005, ML120450843 NCP-2012-001 78 37

Appendix - Comparison of 1967 GDC 2 with 10 CFR 50, Appendix A, GDC 2 1967 GDC Criterion 2, 1967 - Performance Appendix A to Part 50, General Design Standards (Category A) Criteria for Nuclear Power Plants, Criterion 2Design bases for protection against natural phenomena.

Those systems and components of reactor Structures, systems, and components facilities that are essential to the prevention important to safety shall be designed to of accidents which could affect the public withstand the effects of natural health and safety, or to mitigation of their phenomena such as earthquakes, consequences, shall be designed, tornadoes, hurricanes, floods, tsunami, fabricated, and erected to performance and seiches without loss of capability to standards that will enable the facility to perform their safety functions. The withstand, without loss of the capability to design bases for these structures, protect the public, the additional forces that systems, and components shall reflect:

might be imposed by natural phenomena (1) Appropriate consideration of the such as earthquakes, tornadoes, flooding most severe of the natural phenomena conditions, winds, ice, and other local site that have been historically reported for effects. The design bases so established the site and surrounding area, with shall reflect (a) appropriate consideration of sufficient margin for the limited the most severe of these natural accuracy, quantity, and period of time in phenomena that have been recorded for which the historical data have been the site and the surrounding area, and (b) accumulated, (2) appropriate an appropriate margin for withstanding combinations of the effects of normal forces greater than those recorded to and accident conditions with the effects reflect uncertainties about the historical of the natural phenomena and (3) the data and their suitability as a basis for importance of the safety functions to be design. performed.

Applicability of 10 CFR 50, Appendix A, GDC 2 to Diablo Canyon PG&E committed to address any exceptions taken to Appendix A to Part 50, General Design Criteria, during the original Diablo Canyon licensing process. 79 Prior to the NRC issuing the Operating License, PG&E stated that the Diablo Canyon conforms to 10 CFR 50, Appendix A, GDC 2, (without exception). 80 The NRC recently issued Notice of Violation (VIO 05000275;323/2012-004-01, Failure to Incorporate Required Information in the Final Safety Analysis Report Update) 81 associated with the failure of PG&E to include this information in the FSARU.

End Notes:

1 Report on the Analysis of the Shoreline Fault Zone, Central Coast California to the USNRC, PG&E , January 2011, Figure 6-19, page 6-51, ADAMS ML110140400 2

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

October 12, 2012 ML120730106 3

Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011005 and 05000323/2011005 (ML120450843), Section 1R15, Operability Evaluations, February 14, 2012 4

Non-Concurrence, NCP-2012-001, Diablo Canyon Power Plant Inspection report 050000275/323-2011005, ML120450843 38

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

October 12, 2012, ML120730106 6

Operation - Safety and Compliance, Part 9900: Technical Guidance http://www.nrc.gov/reading-rm/doc-collections/insp-manual/technical-guidance/tg-operation-safety.pdf 7

RIS 2005-20, NRC Inspection Manual, PART 9900: Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety, Attachment, Section 3.1, Current Licensing Basis (http://pbadupws.nrc.gov/docs/ML0735/ML073531346.pdf) 8 Regulatory Guide 1.186, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, (http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/rg/division-1/division-1-181.html), 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 9

NEI 97.04, Guidance and Examples for Identifying 10 CFR 50.2 Design Bases, Appendix B, page B21, Seismic Topical Design Bases. ML003678532, (https://adamsxt.nrc.gov/WorkplaceXT/IBMgetContent?vsId={D8B2D4B4-E3BD-4488-B75B-E832F3B33F5D}&objectType=document&id={862C2A33-9C8C-44D8-833F-E54E3D7F44A6}&objectStoreName=Main.__.Library) 10 Ibid 9, Page B21 11 PG&E stated that Diablo Canyon conforms to 10 CFR 50, App A, GDC 2, Letter to FJ Miraglia, NRC, Division of Licensing, from PA Crane, PG&E, September 10, 1981 12 10 CFR 50.34 Contents of Applications; Technical Information 13 10 CFR 50.71, Maintenance of Records, Making of Reports, 14 Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e), ML003740112; endorsed use of NEI 98-03, Revision 1,Guidelines For Updating Final Safety Analysis Reports (http://www.nrc.gov/reading-rm/doc-collections/reg-guides/power-reactors/rg/01-181/)

15 The Diablo Canyon CLB designated the following SSCs as Seismic Category I. SSCs listed per RG 1.29 16 Seismic Evaluation for Postulated 7.5M Hosgri Earthquake, DCPP Units 1&2, PG&E 17 Areva Replacement reactor head, Calculation 6 CS 20327, Appendix 2, revision A, Primary Stress Evaluations, Design Conditions DE 3%, DDE 4% + LOCA, HE 4% + Displacement 18 Ibid 17 19 Diablo Canyon, Unit 1, Current Facility Operating License DPR-80, Tech Specs, ML09181008 (http://adamswebsearch2.nrc.gov/webSearch2/doccontent.jsp?doc={B9458677-D714-43C8-A0C0-12DFC3A173EF})

20 Regulatory Issue Summary 2005-20, Revision to NRC Inspection Manual Part 9900 Technical Guidance, Operability Determinations & Functionality Assessments for Resolution of Degraded or Nonconforming Conditions Adverse to Quality or Safety. (http://pbadupws.nrc.gov/docs/ML0734/ML073440103.pdf) 21 Ibid 7, Section 3.8, Operability 22 Ibid 7, Section 3.10, Specified Safety Function 23 Inspection Procedure 71111.15, Operability Determinations and Functionality Assessments, for a two unit site (http://pbadupws.nrc.gov/docs/ML1120/ML112010663.pdf) 24 Event Number 44675, Offsite Notification and Media Briefing due to Potential Discovery of Off Shore Fault near Plant, November 21, 2008 25 Notification 50086062, LTCA-Ident of Seis Lineament Offsiter, November 14, 2008 26 Diablo Canyon Power Plant, Unit Nos. 1 and 2 - NRC Preliminary Review of Potential Shoreline Fault, April 8, 2009 27 Notification 50086062, LTCA-Ident of Seis Lineament Offsiter, November 14, 2008 28 Notification 50341463, NRC SRI Question on the Shoreline Fault Study, September 14, 2010 29 Notification 50086062, LTCA-Ident of Seis Lineament Offsiter, Task 24, October 10, 2010 30 Notification 50086062, LTCA-Ident of Seis Lineament Offsiter, Task 30, December 16, 2010 31 PG&E submitted to the NRC Report on the Analysis of the Shoreline Fault, Central Coast California, January, 7, 2011, ML110140400 32 E-Mail and Attachment, from Michael Peck to Geoffrey Miller and et al,

Subject:

ACT: Diablo Canyon -

Recommendation for Regulatory Disposition, Attachments: Diablo Canyon Seismic White Paper.docx, February 3, 2011 33 Report on the Analysis of the Shoreline Fault Zone, Central Coast California to the USNRC, PG&E , January 2011 34 From Figure 6-19, Report on the Analysis of the Shoreline Fault Zone, Central Coast California to the USNRC, PG&E , January 2011 35 Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011002 and 05000323/2011002, May 11, 2011, (http://adamswebsearch.nrc.gov/idmws/ViewDocByAccession.asp?AccessionNumber=ML111310608) 36 Task Interface Agreement (TIA) - Concurrence on Diablo Canyon Seismic Qualification Current Licensing and Design Basis (TIA 2011-010), August 1, 2011, ML112130665 39

37 PG&E Letter DCL-1 1-097, License Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake," ADAMS ML11312A166 38 PG&E submitted Letter DCL-1 1-124, Standard Review Plan Comparison Tables for License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake, December 6, 2011, ML11342A238 39 Non-Concurrence, NCP-2012-001, Diablo Canyon Power Plant Inspection report 050000275/323-2011005, ML120450843 40 Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011005 and 05000323/2011005, Section 1R15, February 14, 2012 (http://adamswebsearch.nrc.gov/webSearch2/doccontent.jsp?doc={D8DD93EB-2036-4A68-8ADC-39F302FFEAEE})

41 Ibid 40 42 Diablo Canyon Power Plant, Unit Nos. 1 And 2 -NRC Review of Shoreline Fault (TAC NOS. ME5306 and ME5307),

October 12, 2012, ML120730106 43 PG&E Letter DCL-12-1 08, Withdrawal of License Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake," October 25, 2012, ML12300A105 44 Revised Response To Task Interface Agreement - Diablo Canyon Seismic Qualification Current Licensing and Design Basis, TIA 2011-010 (TIA 2012-012) (TAC NOS. ME9840 and ME9841), February 14, 2012, ML12297A199 45 Diablo Canyon Power Plant, Unit Nos. 1 and 2 -NRC Review of Shoreline Fault (TAC NOS. ME5306 and ME5307),

October 12, 2012, ML120730106 46 Regulatory Guide 1.181, Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e),

(http://pbadupws.nrc.gov/docs/ML0037/ML003740112.pdf) 47 NEI 98-03, Revision 1, Guidelines for Updating FSARs, June 1999, ML003779023 48 Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, (http://pbadupws.nrc.gov/docs/ML0037/ML003759710.pdf) 49 NEI 96-07, Guidelines for10 CFR 50.59 Evaluations, ML003636043 50 PG&E Letter DCL-1 1-097, License Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake," ADAMS ML11312A166 51 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, March 12, 2012, ML12056A046 & ML12053A340 52 Ibid 7, Section C-1 Relationship Between the General Design Criteria and the Technical Specifications 53 Ibid 7 54 Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, http://pbadupws.nrc.gov/docs/ML1303/ML13038A102.pdf 55 Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, October 1973, http://pbadupws.nrc.gov/docs/ML0037/ML003740213.pdf 56 Ibid 7, Section C.4, Use of Alternative Analytical Methods in Operability Determinations 57 Ibid 54 58 Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, October 1973, http://pbadupws.nrc.gov/docs/ML0037/ML003740213.pdf 59 Non-Concurrence, NCP-2012-001, Diablo Canyon Power Plant Inspection report 050000275/323-2011005, ML120450843 60 Ibid 7 61 Ibid 7, Section 3.10, 62 PG&E Letter DCL-1 1-097, License Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake," ADAMS ML11312A166 63 PG&E submitted Letter DCL-1 1-124, Standard Review Plan Comparison Tables for License Amendment Request 11-05, Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake, December 6, 2011, ML11342A238 64 As defined in 10 CFR 2.102, 2.107, & 2.108, and NRR Office Instruction LIC-109, Acceptance Review Procedures, Revision 1, ML091810088 65 Discussion with Diablo Canyon NRR PM, January 2012 66 Ibid 7, Section C-4, Use of Alternative Analytical Methods in Operability Determinations 67 Supplement No. 7 to the Safety Evaluation Report By The Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission In The Matter Of Pacific Gas And Electric Company Diablo Canyon Nuclear Power Station, Units 1 And 2 Docket Nos. 50-275 And 50-323, 2.5.2 Seismology 68 NRC Approved Code Cases (exceptions to Code requirements), Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III, (http://pbadupws.nrc.gov/docs/ML1018/ML101800532.pdf) 40

69 Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2011005 and 05000323/2011005, Section 1R15, February 14, 2012 (http://adamswebsearch.nrc.gov/webSearch2/doccontent.jsp?doc={D8DD93EB-2036-4A68-8ADC-39F302FFEAEE})

70 Ibid 7, Section 3.9, Reasonable Expectation 71 Ibid 7, Section 4.6.2, Prompt Determinations 72 Diablo Canyon Power Plant, Unit Nos. 1 And 2 -NRC Review of Shoreline Fault (TAC NOS. ME5306 and ME5307),

October 12, 2012, ML120730106 73 Research Information Letter (RIL) 12-01 "Confirmatory Analysis of Seismic Hazard at the Diablo Canyon Power Plant from the Shoreline Fault Zone" (ADAMS Accession No. ML121230035).

74 Ibid 7, Section C.6. Use of Probabilistic Risk Assessment in Operability Decisions 75 Ibid 70 76 Revised Response To Task Interface Agreement - Diablo Canyon Seismic Qualification Current Licensing and Design Basis, TIA 2011-010 (TIA 2012-012) (TAC NOS. ME9840 and ME9841), February 14, 2012, ML12297A199 77 E-Mail and Attachment, from Michael Peck to Geoffrey Miller and et al,

Subject:

ACT: Diablo Canyon -

Recommendation for Regulatory Disposition, Attachments: Diablo Canyon Seismic White Paper.docx, February 3, 2011 78 Non-Concurrence, NCP-2012-001, Diablo Canyon Power Plant Inspection report 050000275/323-2011005, ML120450843 79 Letter, from A. Giambusso, Director of Licensing, Atomic Energy Commission (AEC), to F.T. Searls, Pacific Gas and Electric, dated August 13, 1973 80 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 81 Diablo Canyon Power Plant - NRC Integrated Inspection Report 05000275/2012004 and 05000323/2012004, November 13, 2012, ML12318A385 41

Document 2 - Memo Establishing Panel September 3, 2013 MEMORANDUM TO: Michael Case - Chair Britt Hill - Member Rudolph Bernhard - Member FROM: Eric J. Leeds, Director /RA/

Office of Nuclear Reactor Regulation

SUBJECT:

AD HOC REVIEW PANEL - DIFFERING PROFESSIONAL OPINION INVOLVING SEISMIC ISSUES AT DIABLO CANYON (DPO-2013-002)

In accordance with Management Directive (MD) 10.159, The NRC Differing Professional Opinions Program, I am appointing you as members of a Differing Professional Opinion (DPO)

Ad Hoc Review Panel (DPO Panel) to review a DPO that was forwarded to me to disposition.

The DPO (Enclosure 1) raises concerns on seismic issues at Diablo Canyon.

I have designated Mike Case chairman of this DPO Panel and Britt Hill as a DPO Panel member. Rudolph Bernhard was proposed by the DPO submitter and serves as the third member of the DPO Panel. In accordance with the guidance included in MD 10.159 and consistent with the DPO Program objectives, I task the DPO Panel to do the following:

Review the DPO submittal to determine if sufficient information has been provided to undertake a detailed review of the issue.

Meet with the submitter, as soon as practicable, to ensure that the DPO Panel understands the submitters concerns and scope of the issues. (Normally within 7 days).

Promptly after the meeting, document the DPO Panels understanding of the submitters concerns, provide the Statement of Concerns (SOC) to the submitter, and request that the submitter review and provide comments, if necessary. (Normally within 7 days).

Maintain the scope of the review to not exceed those issues as defined in the original written DPO and confirmed in the SOC.

Consult with me as necessary to discuss schedule-related issues, the need for technical support (if necessary), or the need for administrative support for the DPO Panels activities.

Perform a detailed review of the issues and conduct any record reviews, interviews, and discussions you deem necessary for a complete, objective, independent, and impartial review. The DPO Panel should re-interview individuals as necessary to clarify information during the review. In particular, the DPO Panel should have periodic

M. Case et al. discussions with the submitter to provide the submitter the opportunity to further clarify the submitters views and to facilitate the exchange of information.

Provide monthly status updates on your activities via email to Renée Pedersen, Differing Views Program Manager (DVPM) about the last day of the month. This information will be reflected in the Milestones and Timeliness Goals for this DPO. Please provide a copy of email status updates to the submitter and to me.

Issue a DPO Panel report, including conclusions and recommendations to me regarding the disposition of the issues presented in the DPO. The report should be a collaborative product and include all DPO Panel members concurrence. Follow the specific processing instructions for DPO documents.

Consult me as soon as you believe that a schedule extension is necessary to disposition the DPO.

Recommend whether the DPO submitter should be recognized if the submitters actions result in significant contributions to the mission of the agency.

Disposition of this DPO should be considered an important and time sensitive activity. The timeliness goal included in the MD for issuing a DPO Decision is 120 calendar days from the day the DPO is accepted for review. The timeliness goal for issuing this DPO Decision is November 29, 2013.

Process Milestones and Timeliness Goals for this DPO are included as Enclosure 2. The timeframes for completing process milestones are identified strictly as goalsa way of working towards reaching the DPO timeliness goal of 120 calendar days. The timeliness goal identified for your DPO task is 70 calendar days.

Although timeliness is an important DPO Program objective, the DPO Program also sets out to ensure that issues receive a thorough and independent review. The overall timeliness goal should be based on the significance and complexity of the issues and the priority of other agency work. Therefore, if you determine that your activity will result in the need for an extension beyond the overall 120-day timeliness goal, please send me an email with the reason for the extension request and a new completion date. I will subsequently forward this request to the DVPM who will forward it to the EDO for approval.

Please ensure that all DPO-related activities are charged to Activity Code ZG0007.

Because this process is not routine, the DVPM will be meeting and communicating with all parties during the process to ensure that everyone understands the process, goals, and responsibilities. The DVPM will be subsequently sending you information intended to aid you in implementing the DPO process.

An important aspect of our internal safety culture includes respect for differing views. As such, you should exercise discretion and treat this matter sensitively. Documents should be distributed on an as-needed basis. In an effort to preserve privacy, minimize the effect on the work unit, and keep the focus on the issues, you should simply refer to the employee as the DPO submitter. Avoid conversations that could be perceived as hallway talk on the issue. We

M. Case et al. need to do everything that we can in order to create an organizational climate that does not chill employees from raising dissenting views.

As a final administrative note, please ensure that all correspondence associated with this case include the DPO number in the subject line, be profiled in accordance with ADAMS template OE-011, be identified as non-public and declared an official agency record when the correspondence is issued. Please email the ADAMS accession number for the record to DPOPM.Resource@nrc.gov and the record will be filed in the applicable DPO case file folder (DPO-2013-002) in the ADAMS Main Library. Following this process will ensure that a complete agency record is generated for the disposition of this DPO. If the submitter requests that the documents included in the DPO Case File be made public when the process is complete, you will be provided specific guidance to support a releasability review.

I appreciate your willingness to serve and your dedication to completing an independent and objective review of this DPO. Successful resolution of the issues is important for NRC and its stakeholders. If you have any questions, you may contact me, Trent Wertz, NRR OCWE Champion, or Renée Pedersen, DVPM, at (301) 415-2742 or email Renee.Pedersen@nrc.gov.

I look forward to receiving your independent review results and recommendations.

Enclosures:

1. DPO-2013-002

2. Milestones and Timeliness Goals cc w/o enclosure: Submitter DVPM

M. Case et al. DPO submitter. Avoid conversations that could be perceived as hallway talk on the issue. We need to do everything that we can in order to create an organizational climate that does not chill employees from raising dissenting views.

As a final administrative note, please ensure that all correspondence associated with this case include the DPO number in the subject line, be profiled in accordance with ADAMS template OE-011, be identified as non-public and declared an official agency record when the correspondence is issued. Please email the ADAMS accession number for the record to DPOPM.Resource@nrc.gov and the record will be filed in the applicable DPO case file folder (DPO-2013-002) in the ADAMS Main Library. Following this process will ensure that a complete agency record is generated for the disposition of this DPO. If the submitter requests that the documents included in the DPO Case File be made public when the process is complete, you will be provided specific guidance to support a releasability review.

I appreciate your willingness to serve and your dedication to completing an independent and objective review of this DPO. Successful resolution of the issues is important for NRC and its stakeholders. If you have any questions, you may contact me, Trent Wertz, NRR OCWE Champion, or Renée Pedersen, DVPM, at (301) 415-2742 or email Renee.Pedersen@nrc.gov.

I look forward to receiving your independent review results and recommendations.

Enclosures:

1. DPO-2013-002

2. Milestones and Timeliness Goals cc w/o enclosure: Submitter DVPM ADAMS Accession Number: ML13242A305 Office NRR NRR Name TWertz ELeeds Date 09/03/13 09/03/13 OFFICIAL RECORD COPY

Document 3 - Panel Report OFFICIAL USE ONLY- SENSITIVE INTERNAL INFORMATION Differing Professional Qpinion :(DPO)

Involving Seismic 'Issues at Diablo Canyon tNuclear :P.ower Plant (DP0-2013-002)

DPO 'P..anel Report

'fiMchaet Case, Panel Chair

..-'" Brittain Hill, *Panel Member Rudolph Bernhard, Panel Memoer OFFICIAL US~ ONLY- SENSITIVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION DPO Panel Report

1. Introduction On July 19, 2013, in accordance with Management Directive (MD) 10.159, 'The NRC Differing Professional Opinions Program," an individual (the submitter) filed a differing professional opinion (DPO) associated with seismic issues at the Diablo Canyon Power Plant (DP0-2013-002), ADAMS Accession No. ML13268A466). By memorandum dated September 3, 2013, the Director of NRR established an Ad Hoc Review Panel (DPO Panel or Panel) in accordance with MD 10.159. Consistent with DPO program objectives, the Director of NRR directed the DPO Panel to conduct a thorough and independent review of the DPO and to issue a report with its conclusions and recommendations.

The issues raised by the DPO submitter occurred over a period from 2010 to the present but generally focused on the agency's consideration of new seismic information related to potential ground motions from the Shoreline and several other earthquake faults near the Diablo Canyon Power Plant (DCPP). Although the DPO Panel focused its review on this issue, during the course of its review, the Panel needed to understand licensee and staff activities that occurred significantly before and after this timeframe. As an example, as part of their consideration, the Panel needed to research staff activities associated with the initial licensing of Diablo Canyon.

The Panel also considered information as current as the agency's response to Fukushima seismic issues. This proved to be an enormous scope of information that ranged across five decades. As appropriate to facilitate understanding of the issue, the DPO Panel has included information it learned about these activities that occurred before and after the specific timeframe associated with the DPO.

2. Background

Diablo Canyon's original seismic evaluations (Design Earthquake [OBE-equivalent for DCPP],

and the Double Design Earthquake [SSE-equivalent for DDPP]) were accepted prior to issuing the Unit 1 Construction Permit in 1968. These seismic evaluations were performed under and met the Atomic Energy Commission's requirements at time of the submittal. For simplicity, the level of peak ground motion (i.e., horizontal or vertical acceleration) expected from an earthquake is commonly expressed as a unit of gravitational acceleration (g, or m/s2 ). The DE/OBE was accepted as being 0.2 g and was thought to be the largest earthquake that was expected to occur during the lifetime of the plant (a 0.2 g earthquake was estimated to occur once in more than 200 years). The ODE/SSE is simply double the ground motion of the largest expected earthquake (DE/OBE), and is not tied directly to any expected earthquake. The higher ground acceleration of the ODE was used to add safety margin to the evaluations and ensure that safety-related structures, systems, and components needed to safely shut the plant down and maintain it safely would function after the earthquake.

In 1973, the licensee for the DCPP, Pacific Gas and Electric Company (PG&E), became aware of the Hosgri fault, which was discovered offshore from the plant during oil exploration. This fault was previously unknown, and no significant earthquake had previously been attributed to an offshore fault in that area. Based on the timing of this new discovery, the NRC was able to include this information in the approval of operating licenses for DCPP (1984 for Unit 1). As part of this approval, the NRC required PG&E to perform a seismic re-evaluation to include the possible effects of the Hosgri fault using the latest NRC requirements (1 0 CFR 100 and Regulatory Guide 1.61 ). The state-of-the-science in seismic evaluation had significantly OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION improved, so the NRC had upgraded its seismic requirements. The NRC obtained assistance in evaluating the Hosgri fault from U.S. Geological Survey (USGS) and other consultants.

When the Hosgri evaluation was completed, the NRC accepted that this fault could possibly produce 0.75 g peak ground acceleration at Diablo Canyon, but such an extreme event was expected to occur once every 2,000 - 25,000 years. Nonetheless, the NRC required PG&E to make substantial plant modifications to be able to withstand 0.75 g and maintain the same level of plant capability as was required under the SSE. The NRC added these site-specific requirements on top of the existing regulatory requirements.

Therefore, DCPP has the following unique licensing aspects:

1. The plant meets NRC's seismic safety requirements through the DE (0.2 g) and ODE (0.4 g) and the Hosgri evaluation (0.75 g).

2. The plant was required and designed to withstand 0.75 g (based on the Hosgri Evaluation) at the same degree of functionality as an SSE.

3. PG&E used two different NRC-approved seismic methodologies that are part of the design and licensing bases for the plant, one for the DE and ODE, and the other for the Hosgri evaluation.

4. The plants were required to have instrumentation installed to cause an automatic reactor trip if onsite seismic sensors register 0.4 g.

5. A license condition was added to require a confirmatory seismic study over the first 10 years of operation using the latest methods to verify that the Hosgri evaluation remained accurate. PG&E completed this one-time action, but has maintained a continuous seismic assessment program, working with USGS and state agencies to maintain state-of-the-science knowledge and further study the region around the plant.

6. PG&E was required to develop a seismic risk assessment.

The operating license included a license condition requiring a confirmatory seismic study over the first 10 years of operation, using the latest methods, to verify that the Hosgri evaluation remained accurate. PG&E completed this one-time action (known as the Long-Term Seismic Program, or LTSP) using both deterministic and probabilistic, state-of-the-science methods.

The LTSP evaluations concluded that the reanalyzed ground motions were generally lower than already considered for the Hosgri evaluation, and that slightly higher ground motions at frequencies >15 Hz were well within the safety margins of the plant. The staff extensively reviewed the study and agreed with its conclusions as documented in a Supplemental Safety Evaluation Report (SSER) in 1991.

The first new significant seismic information to be identified after plant licensing was the USGS reassessment of seismic activity and new indications in 2008 that there may be an offshore fault close to the plant. PG&E reported this information to the NRC and performed an intensive study of what became known as the Shoreline fault. Although some of the physical features of an active fault are not present, others are, so PG&E concluded that a fault was present and reported their study results in January 2011. This study also reevaluated potential ground motions from the Hosgri and other nearby faults. PG&E concluded that potential ground motions from the Shoreline and other reanalyzed faults were lower than previously considered in the LTSP. Therefore, PG&E believed the plant was safe and no further action was needed.

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION The NRC, with contractor support, evaluated potential ground motions from the Shoreline fault based on the licensee's data and independent analyses. The NRC's results for a number of possible cases showed similar but slightly higher results based on some added conservatisms.

Based on these detailed reviews, the NRC issued Research Information Letter 2012-01 to report the evaluation results. The associated cover letter documented the NRC's conclusions that the Shoreline fault report should be treated as a lesser included case under the NRC-approved Hosgri evaluation because the assumptions and calculations appropriately correlated to those used in the Hosgri evaluation. Because the Shoreline results were lower than the Hosgri evaluation (HE) results, this action resolved the inspection question about which set of requirements (DE/ODE, or HE) should be used to assess the safety impact to the plant and the impact to plant safety.

2.1. Use of Seismic Ground Motions in Safety Analyses In simplest terms, earthquakes create waves of energy that travel through the Earth and produce vibratory ground motions at the surface. Like ocean waves, seismic waves can be reflected or refracted as the move away from the source, and can be amplified or muted as they approach the Earth's surface. Seismic ground motions commonly are represented as response spectra, which plot the spectral frequency of vibration against the expected level of ground acceleration (Figure 1).

2.5 ~,---------------,---------, This relationship between the Free Field GMRS frequency of a seismic wave (or 5% damping ground motion) and level of 2 ground acceleration is important.

Different frequencies of ground

-o-DDE. Fig 3.7-4 motion have different effects on

§ 1.5 --HE. Fig 2.5- OQ' different structures, systems, and c 29+30 d '!\

0

,0 Q components. Ground motions

~..

~ 1 I 0.5

,0 '\D DO I

I d

\

it

' ' ... CJ.. ... ___ o from are different compared earthquakes by their often "peak ground accelerations," which is the level of acceleration that occurs with spectral frequencies of 100 Hz. Even though we refer to the DOE has having a "peak 0

_ ground acceleration" of 0.4 g, 0.1 1

'° Frequency (Hz) 10 0 100 0 Figure 1 shows that much larger Figure 1. Characteristic response spectra for Diablo Canyon NPP, ground accelerations are from FSARU figures indicated in legend. considered for spectral frequencies lower than 20 Hz.

Many important structures, systems and components at nuclear power plants are sensitive to these lower frequency accelerations.

For most nuclear power plants, the Safe Shutdown Earthquake (SSE) is represented by a single response spectrum, which represents the maximum vibratory ground motion expected for a site.

A typical response spectrum (like those in Figure 1) is developed for the free-field surface response, meaning that the calculations assume there are no engineered structures resting on the surface.

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION The Diablo Canyon NPP was licensed before the SSE concept was established in regulations.

Instead of a single SSE, seismic analyses for Diablo evaluated the maximum earthquake potential with two distinct earthquakes: the double design earthquake (ODE) and Hosgri earthquake (HE). The DOE is represented by a free-field response spectrum that was developed using standard methods. However, the HE used several different methods in order to account appropriately for the physical characteristics of larger ground motions. Some of the HE analyses developed free-field response spectra (as shown in Figure 1), whereas other HE analyses accounted for the effects of large structures being present at the site. In addition, the HE analyses also accounted for some non-linear effects in the near-surface site response, which was not a significant consideration for the DOE analyses.

In addition to the different analytical methods for the DOE and HE, different assumptions can be made about the amount of energy lost by the ground vibrations due to friction and heating (e.g.,

NRC Regulatory Guide 1.61 ). This energy loss is referred to as "damping," with higher damping values representing larger energy losses (i.e., lower magnitude accelerations). As shown in Regulatory Guide 1.61, larger damping values are acceptable for SSE analyses compared to Operating Basis Earthquake analyses. These higher damping values reflect the larger energy losses expected in larger magnitude ground motions. Typically, a damping value of 5% is used as a reference value, although higher or lower damping values can be used in different safety analyses.

For the Diablo Canyon NPP, higher damping values were used in HE evaluations than for most ODE evaluations. It is important to note that the HE values are consistent with the values recommended for the SSE in Regulatory Guide 1.61. The ODE, which is viewed as equivalent to the SSE, used damping values that generally were lower than allowable for the SSE. In other words, many of the ODE analyses were conservative and allowed for more efficient energy transfer through structures, systems, and components than normally would be assumed in SSE analyses.

Figure 2 illustrates the importance 2.5  ;----;=================================;l Containment Building. Intake Structures and Concrete structures of these differences in analytical methods and assumptions.

2 ; DDE. Fig 3.7-4 Analyses of seismic loads on the

......-HE 7%. Fig 3.7-4N+O Diablo Canyon containment I --HE 5%. Fig 2.5-29+30 building used 5% damped, free-

§ 1.5 field ground motions for the ODE.

l:

0 However, the HE analyses used

!. 7% damped, foundation-filtered "ii

1 ground motions, which are

<(

approximately 10% lower than the DOE ground motions for 7-10Hz 0.5 frequencies. Load calculations on the containment building are sensitive to this frequency range.

0 Consequently, these different 0.1 1

"° Frequency (Hz) 10 0 100.0 methods and assumptions result in Figure 2. Response spectra for Diablo Canyon NPP containment the DOE creating higher calculated building and other sturctures, from FSARU figures indicated in loads on the containment building legend. than the HE. For some other important structures, systems, and OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION components, the DOE also represents higher calculated loads than the HE (e.g., reactor coolant pressure boundary components, FSARU section 5.2.1.15).

Because of the complex development of the Diablo Canyon NPP ground motion analyses, there is no single response spectrum that appropriately represents the level of seismic ground motion that was used in the safety analyses. In order to accurately compare ground motions from new information with the current licensing basis, the ground motions in question need to have:

1) Comparable response surfaces (i.e., free-field versus foundation filtered);

2) Comparable approach to modeling nonlinear effects, if any, and;

3) Comparable damping values, which correspond to the damping used in the specific safety analyses.

NRC reviews of the SAR, LTSP, and Shoreline report, along with additional discussions in IPEEE and Gl-199 evaluations, clearly show that different analytical methods and assumptions can be used acceptably to derive appropriate response spectra. Regardless of analytical approach used, ground motions in the current licensing basis for the Diablo Canyon NPP have potentially significant differences in surface loading, nonlinearity, and damping that must be recognized to compare modeling results accurately.

3. Statements of Concerns On October 23, 2013, the DPO Panel met for the first time with the submitter to discuss his DPO submittal and his perspective on the concerns. Prior to the meeting, the DPO Panel reviewed the DPO submittal and identified seven areas that looked like potential concerns. The Panel provided this to the submitter for his consideration. The submitter narrowed his concerns to the following:

The NRC did not enforce the Diablo Canyon Technical Specifications with respect to this seismic issue, because the new seismic information showed that SSCs could be exposed to greater vibratory motion than previously considered for the SSE PG&E's operability evaluation following the development of the new seismic information was inadequate, because the new seismic information was not compared correctly to the plant's licensing basis.

The NRC failed to enforce 10 CFR 50.59 requirements that PG&E obtain an amendment to their license, because the new seismic information showed that more than a minimal increase would occur in the likelihood of sse malfunction.

The NRC failed to adequately address the Los Osos and San Luis Bay faults, which could produce ground motions in excess of the SSE ground motion.

The full statement of DPO concerns is included as Appendix A and was used by the Panel to focus its activity.

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

4. Evaluation In support of its independent evaluation of DP0-2013-002, the DPO Panel met and communicated with the DPO submitter initially and throughout the DPO process to obtain his perspectives on the concerns as well as to discuss with him the status and results of its review.

The Panel also reviewed the documents, records, and references cited throughout the DPO report and listed in Appendix 8, "Records and Documents Reviewed by the DPO Review Panel." The Panel also interviewed other individuals related to DPO issues to obtain additional background information and the processes that were (or are being) followed by the licensee and the staff to address the issue. The Panel members invested a considerable amount of time reading the extensive record associated with this issue. Finally, the Panel members met among themselves to plan their work, to review the issues, and to document their conclusions and recommendations.

4.1. Factors Framing the Evaluation In order to complete its decisions of the DPO concerns, the Panel needed to weigh and place into a contextual framework a number of issues that relate to the DPO. The Panel sought to develop a basis for a decision on the DPO concerns and not a detailed argument for who is right. Safety concerns were the overriding factor, although other factors contribute significantly to understanding the DPO concerns. These factors and the Panel's underlying understanding of these factors are explained below.

4.1.1. Treatment of New Siting-related Information As part of completion of the license condition for the Long-Term Seismic Program, PG&E committed to "maintain a strong geosciences and engineering staff to keep abreast of new geological, seismic, and seismic engineering information and evaluate it with respect to its significance to Diablo Canyon, ... " (NRC, 1991, p. 2-49). The NRC did not specify exactly how PG&E would evaluate new information. Nevertheless, there was a clear expectation that new seismic information would continue to be evaluated for significance, without the need for the NRC to take additional regulatory actions to initiate such evaluations.

The NRC expects licensees to evaluate new information that has the potential to affect the licensing basis of the plant, based on the applicable regulatory requirement (e.g., operability determinations, 10 CFR 50.59 evaluations, corrective actions under Criterion XVI of 10 CFR 50, Appendix 8 and other quality assurance programs). However, guidance that specifically addresses this issue could be improved, as discussed in the conclusions.

4.1.2. Unique Diablo Canyon Seismic Design Basis The seismic design basis for Diablo Canyon is both the Double Design Earthquake and Hosgri Evaluation. Throughout the FSARU, both the Double Design and Hosgri earthquakes are used to design and qualify SSCs that are important to safety. This basis has been well established from the time of the operating license, through the LTSP evaluation until the current time.

Nevertheless, applicable regulations and review guidance are designed to evaluate a single seismic design basis.

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION 4.1.3. Ambiguity in the FSARU For a variety of reasons (such its role of documenting historical information, writing style, complexity of the seismic design basis, lack of guidance on new information), the FSARU is not always as clear as it could be with respect to the seismic licensing basis and how we use that basis to evaluate new seismic information. For example, one FSARU section implies that only certain SSCs were designated to withstand the Hosgri earthquake. In the Prompt Operability Assessment (POA), the licensee clarified that all seismic Category I SSCs were evaluated for the 1977 HE. Consequently, a reasonable person could easily draw different meanings from the seismic information in the FSARU.

4.1.4. Risk Insights Despite the complexity of the licensing issues, from a risk and safety perspective, the Diablo Canyon NPP is seismically robust. The Diablo Canyon NPP is relatively well-studied from a seismic risk perspective. In 1979, the staff evaluated seismic risk associated with the Diablo Canyon NPP without the "Hosgri fix" and estimated the likelihood of core damage from seismic events to be approximately one chance per 22,000 years. In the 1991 LTSP, a more extensive evaluation was conducted by PG&E and reviewed by the staff. This review included consideration of both the Los Osos and San Luis Bay seismic sources and estimated the core damage frequency from seismic events to be approximately one chance per 27,000 years.

These seismic core-damage results are comparable to other nuclear power plants that were evaluated in the Long-term Seismic Program and as part of the Individual Plant Examination of External Events program in the early 1990's (see NUREG-1742).

4.2. Evaluation of Specific DPO Concerns Concern #1 -The NRC has not enforced Diablo Canyon Technical Specification requirements that key plant safety equipment remain operable during reactor operation.

New seismic information developed by Pacific Gas and Electric concluded that Technical Specification required Structures, Systems and Components (SSCs) can be exposed to greater vibratory motion than was used to qualified this equipment for the facility safe shutdown earthquake (SSE) design basis. For Technical Specification required SSCs to be considered operable, the licensee is required to demonstrate a reasonable assurance that this plant equipment would still be capable of performing the safety functions in accordance with the plant design bases and safety analysis.

The Panel believes that the NRC has properly evaluated the licensee's determination of operability as presented in Prompt Operability Assessment (POA) of October 21, 2011, and as guided by NRC Inspection Manual Part 9900. The requirement for this concern is driven by the plant's Technical Specifications which, in many cases, prescribe direct surveillance requirements. In this specific circumstance, there is not a specific surveillance requirement to demonstrate SSC operability for seismic issues. So the situation of new seismic information on SSCs is assessed against the definition of OPERABLE contained in the facility's Technical Specifications. The definition of operability states:

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety functions, and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication and other auxiliary equipment that are required for the OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION system, subsystem, train, component, or device to perform its function(s) are also capable of performing their related support functions(s).

This definition prescribes the "requirement" for this particular concern and is basically silent on how to accomplish this evaluation. The NRC and its licensees have a long history of precedents in this area and have also developed guidance for this determination (i.e., IMC Part 9900). Neither, however, is a requirement unto itself.

The Panel examined the licensee POA update of October 21, 2011, and believes it to be technically credible and procedurally consistent with IMC Part 9900 guidance. The underlying licensee logic was to compare the ground motions from the new information to previous ground motions where SSC performance has been shown to be adequate. The licensee examined the effect of new information on the DE, DDE, and HE and used insights from the LTSP evaluation.

The DPO submitter had advocated this comparative approach. Although this may not have initially been the case, the updated POA did take all three earthquake levels into account. We agree with the DPO Submitter's approach to the POA, in that examination of all three earthquake levels is appropriate for and consistent with the seismic licensing basis for the plant.

The Panel's evaluation of the technical approach to the operability issue is contained in Section 4.2.1.

In its evaluation of the DDE, the licensee recognized that it was inappropriate to analyze the new ground motions with the "old" DDE calculation methodology. The licensee reassesses the DDE performance using an alternate evaluation methodology, which appears consistent with past approaches used in licensing (e.g., DDE versus HE, LTSP methods). In addition, IMC Part 9900 allows the use of alternate evaluation methodologies in Appendix C.4. The NRC found the use of an alternate methodology to be acceptable in the Shoreline analyses, as alternative approaches were used previously in the FSARU and LTSP to analyze potential ground motions.

The Panel believes that the use of an alternate methodology is technically acceptable and consistent with the NRC operability guidance.

As discussed in section 4.2.1, in March 2014, PG&E developed additional information to allow direct comparison of the ground motions in the 2011 Shoreline report to those used in the FSARU to design and license the plant. This information confirmed the conclusions of the POA.

The POA also recognized that the use of alternate methodologies is only acceptable for operability and not for full compliance with the CLB. This issue is being tracked as a corrective action to close the POA.

Ultimately, the Panel believes that the licensee's expected response to the Fukushima 2.1 seismic issue should provide the appropriate framework for evaluating the potential significance of new seismic information. Re-evaluated ground motions need to be placed into an integrated context with all other seismic and safety information relevant to Diablo Canyon. The Fukushima 2.1 response activity is expected to do that. Although Diablo Canyon has the advantage of a detailed post-licensing evaluation of SSC seismic performance from the LTSP, the methods used in the LTSP are not always current. If the reevaluated seismic hazard for Diablo Canyon turns out to be greater than the plant's design basis, a seismic risk assessment (if warranted) should provide an up-to-date analysis of how SSCs are expected to function at a potentially higher level of seismic hazard.

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION 4.2.1. Technical Assessment of the Potential for Seismic Loads on SSCs to Exceed Previously Analyzed Conditions The crux of the DPO submitter's concern focuses on a potentially important safety consideration: do new ground motions in the 2011 Shoreline report (PG&E, 2011) exceed the levels of ground motion considered in the FSARU for design and qualification of Category 1 SSCs? The DPO submitter asserts that this exceedance occurs, and that the licensee and NRC should have taken additional actions to ensure SSG operability. However, if the new ground motions were actually lower than those already used in the FSARU to design and license the plant, then further assessments would not be warranted.

For this concern, the Panel determined that the evaluations to-date may not have fully considered the potential significance of the new ground motions on the existing FSARU licensing basis for Diablo Canyon. The DPO submitter identifies some shortcomings in previous evaluations, but also makes incorrect comparisons between the new information and information in the FSARU to reach a conclusion about operability and appropriate licensee and NRC actions. The incorrect comparisons appeared to have occurred, however, because PG&E provided insufficient information in the 2011 Shoreline report to appropriately compare the new ground motions to the range of ground motions actually used in the seismic analyses described in the FSARU. Thus, additional information was needed by NRC, PG&E, and the DPO submitter to make correct comparisons.

In the 2011 Shoreline report,

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§ 1.5 evaluation and the Long-Term Seismic Program. This comparison only evaluates results for a standard reference condition, which is a free-field ground motion with 5% damping (Figure 3). As 0.5 shown in Figure 3, if the original Hosgri evaluation represented the 0

[.____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___) largest ground motions actually 0.1 1.0 10.0 1oo.o considered in the design and Frequency (Hz) qualification of SSCs, then the f:Tgure3Grou-nd-motion-5 from indicated figures-ril-FSARu-rooE,-- newer ground motions would HE), Long-term seismic program (L TSP), and 2011 PG&E Shoreline clearly be lower than already report (ergodic method).

considered in the original FSARU licensing basis. However, as discussed in the background information (see Figure 2), some SSC analyses used ground motions for the Hosgri evaluation that were effectively lower than DDE ground motions. This situation occurs because most safety analyses were done with ground motions that were different than the 5% damped free-field reference condition.

Returning to the containment building example from Figure 2, seismic loads for the Hosgri evaluation were represented by a 7% damped, foundation-filtered ground motion. In contrast, DDE seismic loads in these analyses were calculated with a 5% damped, free-field ground OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION motion. Thus, the ODE created the highest accelerations (i.e., structural loads) at spectral frequencies of 7 to 11 Hz. If these DOE and HE ground motions were inappropriately compared to the new ground motions shown in the 2011 Shoreline report (Figure 4), it would appear that the new ground motions might exceed the levels previously considered in the FSARU for frequencies of 5 to 30 Hz. Nevertheless, the assumed validity of this incorrect comparison is a fundamental assumption for the logic in the DPO submittal regarding the need for additional actions by both PG&E and NRC.

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~ 0.8 ground motions used to design 0.6 and license the plant, additional information was needed from 0.4 PG&E. This information would need to consider if other levels of 0.2 L . . . - - - - - ' - - - - - - ' - - - - - - - - - - - - - ' damping should be used for the 1.0 10.0 100.0 I Frequency (Hz) new ground motions, such as

-- I those corresponding to a Safe Figure 4. Ground motions from FSARU and 2011 Shoreline Report figures (ergodic method), incorrectly assuming the new ground Shutdown Earthquake in motions are directly comparable to FSARU inputs for containment Regulatory Guide 1.61. PG&E building analyses. also would need to consider if other potentially significant effects, such as the presence of building foundations or non-linear material responses (e.g., FSARU rev 21, section 2.5.3.1 0), should be considered for the new ground motions. These considerations are not expressed in the 2011 Shoreline report, or in staff's previous evaluations of the issues surrounding the Shoreline Fault ground motions, or in the DPO submittal.

On 19 December 2013, Panel members discussed this issue of ground-motion comparability with PG&E staff, and outlined the need to compare the new ground motions with the ground motions actually used in the FSARU analyses for design and qualification of safety-related SSCs. PG&E agreed to conduct additional analyses of the new ground motions, so that the results of these analyses would be directly comparable to the inputs used in the FSARU analyses rather than an alternative metric such as the LTSP.

On 5 March 2014, Panel members reviewed additional calculations that were developed by PG&E to allow for direct comparison of potential ground-motions in the 2011 Shoreline report to the ground motions used in the FSARU analyses. PG&E calculated in-structure acceleration response spectra as the basis for comparison, as these spectra already were available for the ODE and HE from FSARU section 3.7 analyses.

To convert the 2011 Shoreline, Los Osos, and San Luis Bay ground-motion spectra to in-structure acceleration response spectra, PG&E developed a scaling relationship from the LTSP analyses that compares the calculated free-surface ground motion to an in-structure response OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION spectrum. This scaling relationship accounts for the effects of processes such as soil-structure interaction and the presence of building foundations. PG&E applied this scaling factor to the 2011 Shoreline, Los Osos, and San Luis Bay ground-motion spectra to calculate in-structure response spectra for 5% damping. PG&E used both ergodic and single-station ground motions from the 2011 Shoreline report.

To account for the different damping values used to analyze the seismic performance of different SSCs (i.e., FSARU rev 21, section 3. 7.1.3), PG&E used analytical methods in Pacific Earthquake Engineering Research Center report 2012/01 (Spectral Damping Scaling Factors for Shallow Crustal Earthquakes in Active Tectonic Regions) to develop scaling factors. PG&E applied these scaling factors to the 5% damped in-structure response spectra for the Shoreline, Los Osos, and San Luis Bay faults (SLS), to develop response spectra for the different damping values shown in Table 1. Although most of the damping values used for these faults correspond to SSE values in NRC Regulatory Guide 1.61, PG&E used slightly lower damping values (i.e., more conservative) in several analyses.

Table 1. Scaling factors used in March 2014 PG&E analyses.

Percentage Damping Type of SSC ODE HE SLS Containment structures 5 7 7 Welded structural steel assemblies 1 4 4 Bolted or riveted steel assemblies 2 7 7 Mechanical components 2 4 3 Vital piping systems (except RCL) >12" 0.5 3 3 Vital piping systems (except RCL) <12" 0.5 2 2 Reactor Coolant Loop 1 4 3 Steam Generators 4 4 3 Integrated Head Assembly 6.85 6.85 6.85 Control Rod Drive Mechanisms 5 5 5 For each of the 10 classes of SSCs (i.e., FSARU rev 21, section 3.7.1.3), PG&E first plotted frequency versus acceleration response for the highest values from either the DOE or HE analyses. PG&E then compared the appropriately scaled Shoreline, Los Osos, and San Luis Bay in-structure response spectra to the DDE+HE spectrum. These comparisons used both the ergodic and single station results from the 2011 Shoreline report.

The in-structure response spectra for the reanalyzed Shoreline, Los Osos, and San Luis Bay faults were all lower than the DDE+HE response spectrum, for both ergodic and single-station results at spectral frequencies of <30 Hz. For several SSCs, the ergodic response spectra met or slightly (<10%) exceeded the DDE+HE spectrum at spectral frequencies of 30-50Hz. This small high-frequency exceedance would not be expected to significantly affect the performance of these types of SSCs. In addition, most of the slight exceedances occurred for SSCs that PG&E had selected a conservative damping value (i.e., lower than used for HE analyses). All of the reanalyzed single-station response spectra were lower than the DDE+HE response spectrum.

In summary, PG&E reanalyzed the ground motions from the 2011 Shoreline report using the same assumptions as in the FSARU for damping level and foundation filtering. The reanalysis allows for direct comparison of the in-structure responses from potential earthquakes on the Shoreline, Los Osos, and San Luis Bay faults to the in-structure responses that were used to OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION design and license the plant. Nearly all the reanalyzed in-structure response are lower than those used to design and license the plant, with the exception of slight (<10%) exceedances at 30-50 Hz spectral frequencies for several SSCs using ergodic analyses. These slight exceedances arise, in large part, from conservative damping values used by PG&E and are not judged significant for the SSCs being considered. The Panel concludes that these comparisons are appropriate, and that potential ground motions from faults characterized in the 2011 Shoreline report do not exceed the levels of in-structure acceleration already considered in the design and licensing of the plant.

4.2.2. Summary of Concern #1 To summarize the Panel's assessment of Concern #1, the DPO raised an important issue that highlights the complexity of information used to assess the seismic loads on safety related SSCs during the licensing and construction of Diablo Canyon NPP. Nevertheless, the DPO inappropriately compares different types of ground motions to incorrectly conclude that sse functionality should be re-assessed, and asserts that NRC staff did not respond appropriately to new information. This mis-comparison appears understandable, as appropriate ground-motion information was not available to NRC, PG&E, or the DPO submitter to make a correct comparison.

Previously, NRC and PG&E staffs reached an apparently reasonable conclusion that the new ground motions were bound by existing ground motions (i.e., the Hosgri evaluation). Thus, no further analyses appeared warranted, and staff's approach on additional licensing or enforcement actions appears defensible. Based on the Panel's current understanding, this conclusion only appears supportable when all the ground motions are compared to a common reference condition of 5% damping, free-field response. However, most of the Diablo Canyon safety analyses were not conducted at this reference condition. The FSARU analyses used two different ground motions, each of which used different damping values and, at times, different analytical assumptions, which do not always correspond to the common reference condition used in the 2011 Shoreline report. As a result, only a few of the ground motions in the 2011 Shoreline report are directly comparable to the actual ground motions used in the FSARU safety analyses.

In the previous analyses, neither PG&E nor NRC staff, nor the DPO submitter, appeared to recognize the need to compare the new information more clearly to the licensing basis in the FSARU. The need for this comparison apparently was not identified because of the complex differences between the reference ground motion conditions and the range of conditions actually considered in the FSARU analyses. Nevertheless, the DPO submitter succeeded in raising awareness of these important differences, and illustrating how seemingly reasonable interpretations resulted in different implications for operability and safety. As discussed extensively in section 4.2.1, the Panel concludes that once appropriate comparisons are made, potential ground motions from faults characterized in the 2011 Shoreline report do not exceed the levels of in-structure acceleration already considered in the design and licensing of the Diablo Canyon Nuclear Generating Station.

Concern #2 - Pacific Gas and Electric's operability evaluation following development of the new seismic information was inadequate. Comparison of the new seismic information only against the Hosgri Event (HE) and Long Term Seismic Program (LTSP) ground motions was not adequate to demonstrate Technical Specification required SSCs were operable. Neither the HE not the LTSP methods were approved to be used in SSE safety analysis. The HE and LTSP methods over-predicted SSC performance when OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION compared to the SSE design basis methods. Even though the HE and LTSP include higher ground motions, neither of these methods were bounding for plant Technical Specification SSCs seismic qualification. Use of the HE and LTSP ground motions failed to demonstrate that the requirements of the American Society of Mechanical Engineers' (ASME) Boiler and Pressure Vessel Code acceptance limits would be met at the higher ground motions. 10 CFR 50.55a required that ASME acceptance limits be met for plant safety Class 1 and 2 following an SSE. Demonstration that the ASME acceptance limits are met provides assurance that the integrity of key plant systems, including the reactor coolant pressure boundary would be maintained following the higher seismic stress levels represented by the new seismic information.

The Panel does not believe that Concern #2 raises new fundamental issues with respect to the seismic safety issue that is not already discussed in the Panel's consideration of Concern #1.

However, the concern raises some considerations on evaluation methods and the ASME Code that the Panel addressed below.

During plant operation, conditions or equipment changes that are outside what is considered normal can occur. For failures associated with the Technical Specification's requirements, specific testing to determine equipment operability is often provided and Action Statements are used for the timing of actions due to the condition under construction. For conditions that are not as well defined, equipment inoperability is determined to exist at the time there is sufficient evidence that the equipment is not capable of meeting its design basis function.

For situations without specific technical specification testing requirements, evaluations can be performed by the licensee to determfne 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.

Concern #2 suggests that there is only one appropriate evaluation method in this case, which is to substitute new seismic information into the original DOE method. In the Panel's estimation, there were three viable evaluation methods to assess seismic performance of plant equipment in the DPO scenario. The first would be to directly substitute the new information into the calculation construct of the HE and DOE. Although this method would provide the most direct comparison to the FSAR commitments, it would offer very little insight as to how the SSCs would actually perform to seismic loads shown in the new information. This is because the older analytical techniques are overly conservative. Even by 1981 when* the staff issued its SSER supplement, the staff allowed the use of more modern insights (e.g., damping values) because the use of these more conservative DDE values was no longer technically justified.

A second evaluation method available to the licensee would be to use completely up-to-date probabilistic methods. This approach would be similar to the approach used for Fukushima 2.1.

Although this evaluation method would be the most technically credible, it would take a considerable amount of time to complete. Such an approach would not have been responsive enough for the purpose of an operability evaluation.

The final possible evaluation method is the one used by the licensee. This evaluation method involved comparison against the HE and the LTSP. This evaluation method is attractive OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION because the methods used in the LTSP are improved over those of initial licensing. The LTSP was extensively reviewed by the staff and provides an additional regulatory perspective that the staff agreed with the licensee's conclusions in that report that adequate seismic margin is provided in the Diablo Canyon design of SSCs. The NRC staff also thought the LTSP evaluation bound the seismic hazards for the Diablo Canyon NPP. The shortcoming of this evaluation method is that it does not compare directly with parts of the FSARU licensing basis.

The Panel further reviewed this issue as detailed in Concern #1.

As discussed earlier, there is no regulatory requirement known to the Panel that dictates that the only option for evaluating new information is to substitute it into the original licensing basis calculations. Further, the staffs operability guidance specifically allows the use of alternate evaluation methods. Inspection Manual Chapter 0326 provides some insight on operability and functionality in section 03.09 on Reasonable Expectation when it writes:

The discovery of a degraded or nonconforming condition may call the operability of one or more SSCs into question. A subsequent determination of operability should be based on the licensee's "reasonable expectation," from the evidence collected, that the SSCs are operable and that the operability determination will support that expectation. Reasonable expectation does not mean absolute assurance that the SSCs are operable. The SSCs may be considered operable when there is evidence that the possibility of failure of an sse has increased, but not to the point of eroding confidence in the reasonable expectation that the SSC remains operable.

The supporting basis for the reasonable expectation of SSC operability should provide a high degree of confidence that the SSCs remain operable.

The Panel believes that the licensee's method of evaluation meets this standard.

The Panel also sought pertinent guidance to help it understand the potential weakness in the licensee's evaluation approach (i.e. incomplete mapping to the FSARU methods). Guidance to the staff on the balance between safety and compliance during evaluation of plant operations is covered in the Inspection Manual Technical Guidance section. It also indicates that discretion can be exercised in cases where conditions do not pose undue risk. The guidance states, in part:

The NRC has the authority to exercise discretion to permit continued operations-despite the existence of a noncompliance-where the noncompliance is not significant from a risk perspective and does not, in the particular circumstances, pose an undue risk to public health and safety. When non-compliances occur, the NRC must evaluate the degree of risk posed by that non-compliance to determine if specific immediate action is required. Where needed to ensure adequate protection of public health and safety, the NRC may demand immediate licensee action, up to and including a shutdown or cessation of licensed activities.

In addition, in determining the appropriate action to be taken, the NRC must evaluate the non-compliance both in terms of its direct safety and regulatory significance... Based on the NRC's evaluation, the appropriate action could include refraining from taking any action, taking specific enforcement action, issuing orders, or providing input to other regulatory actions or assessments, such as increased oversight (e.g., increased inspection).

Where requirements exist that the NRC concludes have no safety benefit, the NRC can and should take action, as appropriate, to modify or remove such requirements from the OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION regulations or licenses. Requirements that are duplicative, unnecessary, or unnecessarily burdensome can actually have a negative safety impact. They also can tend to create an inappropriate NRC and licensee focus on "safety versus compliance" debates. As the Commission states in its principles of Good Regulation, "There should be a clear nexus between regulations and agency goals and objectives, whether explicitly or implicitly stated."

The Panel believes that by linking the evaluation to the LTSP, the licensee established an important insight to a well-studied (by the staff) seismic risk assessment. The seismic risk of core damage in that study was relatively low (3.7 x 10-5/reactor-year). This level of risk is well below a level that would indicate an immediate safety concern as discussed in LIC-504, "Integrated Risk-Informed Decision Making Process for Emergent Issues." In addition, the letter from the NRC to the licensee on the results of its review of the new seismic information and the staff's 50.54(f) letter on Fukushima 2.1 provide an adequate regulatory footprint to follow up on potential FSARU non-compliances.

Finally, Concern #2 raises issues with respect to 10 CFR 50.55a. The Panel sees no unique issues with PG&E's operability evaluation with respect to 50.55a issues that were not more fully explored in relation to Concern #1. Therefore, the Panel concludes that the associated operability assessment was adequate. 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 Panel's conclusion that both these limits are, at times, applicable as the limiting load. Nevertheless, the relatively low level of damping used in the DOE analyses (e.g., Table 1) results in the DOE creating the limiting load for these SSCs, which is not exceeded by the reanalyzed ground motions from the 2011 Shoreline report (see discussion in Concern #1 ).

The new information by itself did not alter the FSARU approach to maintain both the ODE and HE as faulted conditions with respect seismic component and piping analysis. The Panel's evaluation of Concern #1 concluded that the new information is bounded by the existing DOE and Hosgri evaluations. So the current FSARU conclusions with respect to the ASME acceptance criteria appear valid. In this particular section, the Panel could not see an adequate technical justification to use the new information as a more limiting requirement than those previously identified.

In summary, the Panel believes that in the context of Concern #2, the method used by the licensee was appropriate, with accepted assumptions, to verify that the new information did not indicate the presence of a hazard that would constitute an undue risk to public health and safety. Showing the new postulated hazard is less limiting than already evaluated hazards, using accepted methodology, was one acceptable method for performing the operability evaluation.

Concern #3 -The NRC has failed to enforce the 10 CFR 50.59 requirements that Pacific Gas and Electric obtain an amendment to the Diablo Canyon Operating License prior to incorporating the Shoreline scenario into the FSARU. A license amendment was required because the change resulted in more than a minimal increase in the likelihood of a malfunction of SSCs important to safety than previously evaluated in the FSARU. A license amendment was also required because this change represents a departure from the FSARU method of evaluation used to establish the seismic SSE design basis. The NRC conclusion that a "reasonable assurance of safety" existed was not an adequate OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION basis to conclude an amendment to the Diablo Canyon Operating License was not required.

The DPO Panel believes that the staff did not fail to enforce 10 CFR 50.59 requirements with respect to a proposed update to FSARU Revision 20, in which PG&E was requested to add information concerning the 2011 Shoreline Fault Report. Analysis of this concern requires an understanding of the context of proposed FSARU change itself, the context of where it was placed, and the context of how it was written.

First, this update to FSARU Revision 20 was specifically requested by the staff in a letter documenting the staff's review of the Shoreline Fault. Specifically the staff's letter states:

Therefore, the staff has concluded that the Shoreline scenario should be considered 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).

This created a problematic situation for the licensee because NRC guidelines for FSAR updates suggest that an update might not be warranted. The guidance states that for analyses of new safety issues, the evaluations must be reflected in the FSAR only if, on the basis of the results of the requested analyses or evaluation, the licensee determines that the existing design basis, safety analyses or FSAR description are either not accurate or not binding or both.

Nevertheless, the Panel believes that the FSAR update was appropriate because of the long and at times complex evolution of seismic information for Diablo Canyon. However, the change was likely not required at all, let alone, something that required a license amendment.

The second contextual factor concerns where the updated information was placed. The FSAR update (Revision 21) was placed in the section of the FSAR (Section 2.5) that discusses geology and seismology. The context of the information is that it factually presents results of seismic and geological information about the site, and provides additional explanations of the historical development of the seismic hazards analyses for Diablo Canyon NPP. The FSARU Revision 21 information on the Shoreline Fault zone now discusses the results of both the PG&E and NRC evaluations. The information presented focuses on conclusions from several seismic and geological investigations, which generated little controversy in the DPO submittal.

However, the update did not embellish the description with respect to how the conclusions are used in seismic design, which is an area of DPO controversy. A plain reading of FSAR Revision 21 would indicate that the update has little or no direct 50.59 implications.

Finally, as the DPO submitter suggests, it may be appropriate to consider the implications of the use of this new information with respect to 50.59. As noted earlier, the writing style of FSAR Revision 21 in this section is factual and historical. Although it did include a reference to the NRC letter on the review of the Shoreline Fault, the FSAR update did not include important contextual information from the NRC letter. The first piece of contextual information is that the NRC evaluation was preliminary. Second, PG&E is required to take action consistent with the staff's 10 CFR 50.54(f) letter on Fukushima seismic issues and that "changes to the licensing basis may be appropriate to capture the information developed in response" to the Fukushima seismic issue. Finally, after the Fukushima seismic letter was issued, PG&E committed to providing NRC with an interim evaluation if new information is uncovered that would suggest the Shoreline fault is more capable than currently believed (PG&E, 2012, ADAMS Accession No. ML12300A105). Any such interim evaluation would occur before completion of the evaluations requested in the Fukushima seismic letter. Although the Panel believes that the information OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION requested in the Fukushima seismic letter should provide a comprehensive basis to evaluate potential seismic hazards at the Diablo Canyon NPP, incorporating current information on the Shoreline fault into FSARU Revision 21 appears reasonable given PG&E and NRC communications.

Given that contextual information, the DPO Panel assessed the 50.59 evaluation criteria as described in the staff-endorsed NEI report 96-07. The guidance suggests that changes in design requirements for earthquakes should be best treated as potentially affecting the likelihood of a malfunction rather than frequency of occurrence of an accident. Based on the documented information, both PG&E and NRC had concluded that the ground motions from the 2011 Shoreline report were bounded by existing analyses. Thus, the newer ground motions would not be expected to increase the likelihood of malfunction of SSCs that are important to safety. As discussed in DPO Concern #1, incomplete information was available to make appropriate comparisons between the newer ground motions and the range of ground motions used to assess the safety of the Diablo Canyon NPP. The DPO submitter uses the previously available information to conclude that the newer ground motions exceed the plant's design basis and, thus, indicate an increase in the likelihood of equipment malfunction. For the reasons discussed in DPO Concern #1, these ground motions are not directly comparable.

Consequently, there was insufficient basis to conclude that a license amendment was required to address the 2011 Shoreline report, and NRC staff's recommendation for an FSAR update was reasonable.

The DPO Panel evaluated DPO Concerns #1 and #2, and considered the new ground-motion information provided by PG&E to supplement the 2011 Shoreline report. The Panel believes that there is a sufficient basis to conclude that the likelihood of a malfunction has not increased more than minimally (or more specifically as stated in the guidance, "the uncertainties in determining whether a change in likelihood has actually changed (i.e., there is no clear trend towards increasing the likelihood)." Therefore, the Panel believes that an amendment to the operating licensee is not required, and that the FSARU Revision 21 update is an appropriate action in response to the new information.

Concern #4- The NRC failed to adequately address the Los Osos and San Luis Bay faults. The new seismic information concluded that these faults were also capable of producing ground motions in excess of the current plant SSE design basis.

Although this DPO concern has many important similarities to DPO Concern #1, there is an important distinction that warrants clarification. The focus of the 2011 Shoreline report was on assessing the potential significance of the Shoreline fault, which was a newly characterized fault system. In contrast, both the Los Osos and San Luis Bay faults were recognized previously and evaluated as part of the LTSP. Ground motions for these two faults were simply reevaluated in the 2011 Shoreline report with the same updated methods used to assess the Shoreline fault.

As shown in Figure 3 of the Panel report, the reevaluated potential ground motions for the Los Osos and San Luis Bay faults are approximately 10% higher than potential Shoreline fault ground motions.

In the LTSP, ground motions from the Los Osos and San Luis Bay faults were shown to be significantly lower than the ground motions for the Hosgri fault, which was thought to be the bounding ground motion for the Diablo Canyon NPP. Although the Los Osos and San Luis Bay faults were not addressed explicitly in NRC staff's 2009-2011 evaluations, the detailed 2012 NRC Research Information Letter 12-01 (ADAMS Accession No. ML121230035) evaluated these faults in the context of the Shoreline fault system. In addition, staff used information from OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION the LTSP to conclude in RIL 12-01 that these faults were not capable of producing ground motions that challenged the licensing basis of the plant in a deterministic framework (see RIL 12-01, Chapter 5.1 0). Staff also recognized the non-negligible contribution that these faults might make to a probabilistic assessment (see RIL 12-01, Chapter 6), which would consider both the likelihood and magnitude of potential ground motions.

The Panel agrees with the DPO submitter's concern that NRC staff did not clearly and consistently consider the potential ground motions from the Los Osos and San Luis Bay faults in all reports and actions associated with the 2011 Shoreline Report. From a deterministic perspective, this omission is understandable because the ground motions from these two previously analyzed faults did not increase significantly, and were well within the limits already considered explicitly in LTSP analyses. From a risk perspective, initial analyses showed that individual contributions from these faults to the total seismic hazard were small, and bounded by Hosgri fault ground motions (e.g., RIL 12-01, Chapter 6). Nevertheless, the basis for staffs actions and conclusions sometimes were not clear because the Los Osos and San Luis Bay potential ground motions were not addressed explicitly. The DPO has highlighted the need for more explicit consideration of the Los Osos and San Luis Bay faults in future communications, based on the prominence of these faults in the 2011 Shoreline report.

The remainder of this DPO concern focuses on the same issue of ground-motion comparability that was discussed for DPO Concern #1. As shown in Figure 3, potential ground motions for the Los Osos and San Luis Bay faults can be approximately 10% higher than potential ground motions for the Shoreline fault. Ground motions for these three faults were all calculated with the same methods and assumptions. Thus, equivalent changes in the amount of damping or presence of building foundations should have equivalent changes in the calculated ground motions. In other words, the relative relationships between these three ground motion response spectra should not change significantly. Nevertheless, the Los Osos and San Luis Bay potential ground motions shown in Figure 3 have the same limitation as the Shoreline potential ground motions, in that they are not directly comparable to the full range of ground motions used in the FSARU to license Diablo Canyon. As discussed in Concern #1, these faults were considered explicitly in the March 2014 supplemental analyses by PG&E. The reanalyzed ground motions for the Los Osos and San Luis Bay (and the Shoreline) faults do not exceed the level of ground motion already used to design and license the plant.

5. Conclusions Based on the preceding evaluation, the DPO Panel concludes:

1) The review of the DPO circumstances and information did not reveal a significant or immediate concern with the current understanding of seismic safety of the Diablo Canyon NPP.

2) The seismic licensing history at the Diablo Canyon NPP is long, complex and unique, and has been thoroughly evaluated by both the staff and licensee. Unlike other operating plants, seismic safety at the Diablo Canyon NPP has been evaluated using large ground motions from two different earthquakes. However, the safety analyses often use different physical conditions and analytical assumptions for each earthquake. As a result of these differences, PG&E and NRC staffs, and the DPO submitter, were unable to make an appropriate range of comparisons between the plant's licensing basis and new seismic information.

3) The DPO submitter was a positive contributor to both the licensee's and the staff's actions on seismic safety at the Diablo Canyon NPP, especially with bringing attention to important OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION safety relationships that were not always clear in the FSARU or supporting documents. The staff and licensee rationale in this area could have been improved by having a more direct comparison of the new information with the existing seismic licensing basis. As a result of this DPO, additional information was developed by PG&E to clearly demonstrate that potential ground motions from the Shoreline, Los Osos, and San Luis Bay faults would not exceed the levels of ground motion already considered during the design and licensing of the plant.

4) The staff followed its processes for technical specification operability of plant equipment and 10 CFR 50.59 evaluations with a reasonable technical and safety rationale. The staffs Fukushima 2.1 evaluation process is expected to provide an up-to-date assessment of both Diablo Canyon's seismic safety and the staff's evaluations regarding the Shoreline Fault.

5) The lack of a formal regulatory guidance for evaluating new information on natural hazards appears to be a contributing cause in creating many of the differing interpretations for the potential significance of this information.

6. Recommendations

1) Continue the Fukushima 2.1 evaluation process to both confirm the staff's analyses of the Shoreline Fault and assess the continued safe operation of the Diablo Canyon in consideration of the reevaluated seismic hazards at the site.

2) Better define (perhaps through Fukushima 2.2 or other durable regulatory products) the staff position on assessing new information about potential natural hazards at a site.

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OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION Appendix A:

Statement of Technical Concerns, Derived from Diablo Canyon DP0-2013-002

1) The NRC has not enforced Diablo Canyon Technical Specification requirements that key plant safety equipment remain operable during reactor operation. New seismic information developed by Pacific Gas and Electric concluded that Technical Specification required Structures, Systems and Components (SSCs) can be exposed to greater vibratory motion than was used to qualified this equipment for the facility safe shutdown earthquake (SSE) design basis. For Technical Specification required SSCs to be considered operable, the licensee is required to demonstrate a reasonable assurance that this plant equipment would still be capable of performing the safety functions in accordance with the plant design bases and safety analysis.

2) Pacific Gas and Electric's operability evaluation following development of the new seismic information was inadequate. Comparison of the new seismic information only against the Hosgri Event (HE) and Long Term Seismic Program (LTSP) ground motions was not adequate to demonstrate Technical Specification required SSCs were operable. Neither the HE nor the LTSP methods were approved to be used in SSE safety analysis. The HE and LTSP methods over-predicted SSG performance when compared to the SSE design basis methods. Even though the HE and LTSP include higher ground motions, neither of these methods were bounding for plant Technical Specification SSCs seismic qualification. Use of the HE and LTSP ground motions failed to demonstrate that that the requirements of the American Society of Mechanical Engineers' (ASME) Boiler and Pressure Vessel Code acceptance limits would be met at the higher ground motions. 10 CFR 50.55a required that ASME acceptance limits be met for plant safety Class 1 and 2 following an SSE. Demonstration that the ASME acceptance limits are met provides assurance that the integrity of key plant systems, including the reactor coolant pressure boundary would be maintained following the higher seismic stress levels represented by the new seismic information.

3) The NRC has failed to enforce the 10 CFR 50.59 requirements that Pacific Gas and Electric obtain an amendment to the Diablo Canyon Operating License prior to incorporating the Shoreline scenario into the FSARU. A license amendment was required because the change resulted in more than a minimal increase in the likelihood of a malfunction of SSCs important to safety than previously evaluated in the FSARU. A license amendment was also required because this change represents a departure from the FSARU method of evaluation used to establish the seismic SSE design basis. The NRC conclusion that a "reasonable assurance of safety" existed was not an adequate basis to conclude an amendment to the Diablo Canyon Operating License was not required.

4) The NRC failed to adequately address the Los Osos and San Luis Bay faults. The new seismic information concluded that these faults were also capable of producing ground motions in excess of the current plant SSE design basis.

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OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION Appendix 8:

Publically Available Records and Documents Reviewed by DPO Panel

1. Diablo Canyon Power Plant Units 1 & 2 FSAR Update, Revision 21, September 2013.

ML13280A390.

2. Criterion 2, Design Basis Protection Against Natural Phenomena, of Appendix A, "General Design Criteria for Nuclear Power Plants," to Title 10 of the Code of Federal Regulations (10 CFR), Part 50.

3. NEI 96-07, "Guidelines for 10 CFR 50.59 Evaluations," Revision 1, February 2000.

ML003686043.

4. Union of Concerned Scientists, "Seismic Shift- Diablo Canyon Literally and Figuratively on Shaky Ground," November 2013.

5. Rezaeian, S., and others, "Spectral Damping Scaling Factors for Shallow Crustal Earthquakes in Active Tectonic Regions," Pacific Earthquake Engineering Research Center Report 2012/01, July 2012.

6. Licensee Amendment Request 11-05, "Evaluation Process for New Seismic Information and Clarifying the Diablo Canyon Power Plant Safe Shutdown Earthquake," October 2011. ML11298A247.

7. Letter from Barry S. Allen to Nuclear Regulatory Commission, "Withdrawal of License Amendment Request 11-05," October 2012. ML12300A105.

8. Letter from Joseph M. Sebrosky to Edward D. Halpin, "Diablo Canyon Power Plant Units 1 and 2- Withdrawal of an Amendment Request," October 2012. ML12289A076.

9. Letter from James R. Becker to Nuclear Regulatory Commission, "Standard Review Plan Comparison Tables for License Amendment Request 11-05," December 2011.

ML11342A238.

10. Letter from Eric Leeds to All Power Reactor Licensees, "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-lchi Accident,"

February 20, 2014. ML14030A046.

11. Letter from James R. Becker to Nuclear Regulatory Commission, "Report on the Analysis of the Shoreline Fault Zone, Central Coastal California," January 2011.

ML110140400.

12. Memorandum from Kriss M. Kennedy to Robert Nelson, "Task Interface Agreement-Concurrence on Diablo Canyon Seismic Qualification Current Licensing and Design Basis," August 2011. ML112130665.

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

OFFICIAL USE ONLY- SENSITVE INTERNAL INFORMATION

13. Memorandum from Sher Bahadur to Kriss M. Kennedy, "Revised Response to Task Interface Agreement- Diablo Canyon Seismic Qualification Current Licensing and Design Basis, TIA 2011-010 (TIA 202-012)," November 2012. ML12297A199.

14. Letter from Joseph Sebrosky to Edward D. Halpin, "Diablo Canyon Power Plant- NRC Review of Shoreline Fault," October 2012. ML120730106.

15. Research Information Letter 12-01, "Confirmatory Analysis of Seismic Hazard at the Diablo Canyon Power Plant from the Shoreline Fault Zone," September 2012, ML121230035.

16. Memorandum from Brian W. Sheron to Eric J. Leeds, "Research Information Letter 09-001: Preliminary Deterministic Analysis of Seismic Hazard at Diablo Canyon NPP from Newly Identified "Shoreline Fault" April, 2009. ML090330188.

17. Letter from Neil O'Keefe to John T. Conway, "Diablo Canyon Power Plant- NRC Integrated Inspection Report 05000275/2011005 and 05000323/2011005," February 2012. ML120450843.

18. Non-Concurrence Process Record NCP-2012-001, "Diablo Canyon Power Plant-Inspection Report 05000275/323-2011005," June 2012. ML12151A173.

19. Report, "Additional Branch Chief Comments Related to NCP 2012-001 with Annotations," June 2012. ML12284A066.

20. IE Information Notice No. 79-06, "Stress Analysis of Safety Related Piping," March 1979.

ML080310608.

21. Regulatory Guide 1.61, "Damping Values for Seismic Design of Nuclear Power Plants,"

Revision 1, March 2007. ML070260029.

22. Regulatory Guide 1.181, "Content of the Updated Final Safety Analysis Report in Accordance with 10 CFR 50.71(e)," September 1999. ML003740112.

23. Regulatory Guide 1.186, "Guidance and Examples for Identifying 10 CFR 50.2 Design Bases," Revision 0, December 2000. ML003754825.

24. NUREG/CR-1429, "Seismic Review Table," May 1980. ML110880747.

25. NUREG/CR-6919, "Recommendations for Revision of Seismic Damping Values in Regulatory Guide 1.61 ,"November 2006. ML063260342.

26. NUREG-17 42, "Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program, vols. 1 and 2," April2002. ML021270070 and ML021270674.

27. NRC Inspection Manual Part 9900: Technical Guidance, "Operability Determinations &

Functionality Assessments for Resolution of Degraded or Non-Conforming Conditions Adverse to Quality or Safety." April2008. ML051520373.

28. NRR Office Instruction LIC-202, "Procedures for Managing Plant-Specific Backfits and 50.54(f) Information Requests," Revision 2, May 2010. ML092010045.

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OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

29. NRR Office Instruction LIC-504, "Integrated Risk-Informed Decision-Making Process for Emergent Issues," Revision 3, April 2010. ML100541776.

Non-Publically Available ADAMS Records and Documents Reviewed by DPO Panel

1. Memorandum from Bradley W. Jones to John A Grobe, "NRC Sources of Legal Requirements and the Applicability of 10 CFR Part 100 Standards," August 2008, ML082460980 (NONPUBLIC).

2. Memorandum from Michele G. Evans to Eric J. Leeds, "NRC and Licensee Actions in Response to New Information from a Third Party," ML112730055 (NONPUBLIC).

3. NRR Office Instruction LIC-100, "Control of Licensing Bases for Operating Reactors,"

Revision 1, January 2004. ML033530249 (NONPUBLIC).

4. Memorandum from Michael T. Markley to Neil F. O'Keefe, "Response to Senior Resident Inspector Question Regarding the Diablo Canyon Research Information Letter Associated with the Shoreline Fault," October 2012. ML12213A079 (NONPUBLIC).

5. NUREG-0675 Supplement No. 34, Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Plant," June 1991. ML093070113 (NONPUBLIC).

Information not Located in ADAMS and Assumed to be Non-Publically Available

1. Meeting Summary: Pre-Licensing Meeting with PG&E on Plans to Submit a License Amendment to Incorporate Management of New Geotechnical Seismic Information Into Its Design and Licensing Basis, January 2011.

2. Meeting Summary: Pre-Licensing Meeting with PG&E on Proposal License Amendment for a New Seismic and Design Evaluation Process, July 2011.

3. Meeting Summary: Pre-Licensing Meeting with PG&E on Responses to Staff Questions From Previous Public Meeting on January 26, 2011, May 2011.

4. DCPP Form 69-20108, "UFSAR Change Request," June 2013.

5. Prompt Operability Update, "DCPP Shoreline Fault POA 10-21-2011."

6. Memorandum from Meena K. Kanna to Michael T. Markley, "Safety Evaluation DCPP Units 1 & 2 License Amendment Request for Damping Values for the Seismic Design and Analysis of the Reactor Vessel Integrated Head Assembly (IHA)."

7. Memorandum from Catherine E. Kanatas to Edward Williamson, "Legal Process Regarding North Anna Restart Decision," November 2, 2011.

8. Report, "Diablo Canyon Seismic Licensing History Briefly Summarized." November 2013.

9. Report, "Timeline of Seismic Issues at Diablo Canyon." August 2013.

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OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

10. Report, "Resident Inspectors Recommendation for Regulatory Disposition of the Failure of PG&E to Perform on Operability Evaluation Following Discovery of the Shoreline Fault." February 2011.

11. Letter from John F. Stolz to John C. Morrissey, "Staff Evaluation of Probabilistic Seismic Risk Assessment," November 1978.

12. PG&E Letter DCL-88-192 from D.A. Brand to NRC, "Long Term Seismic Program Completion," July 31, 1988.

13. NUREG-0675 Supplement No. 7, "Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Plant," May 1978.

14. NUREG-0675 Supplement No. 8, "Safety Evaluation Report Related to the Operation of Diablo Canyon Nuclear Power Plant," November 1978.

OFFICIAL USE ONLY - SENSITVE INTERNAL INFORMATION

Document 4 - DPO Decision

Document 5 - DPO Appeal Submission NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

Continued Item 11 NRC Form 690 Page 1

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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).

NRC Form 690 Page 2

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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.

NRC Form 690 Page 3

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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.

NRC Form 690 Page 4

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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)

NRC Form 690 Page 5

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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.

41 See footnote 30 NRC Form 690 Page 13

NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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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NRC Form 690 U.S. Nuclear Regulatory Commission Differing Professional Opinion--Appeal (Continued)

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),

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Document 6 - Statement of Views June 27, 2014 MEMORANDUM TO: Mark A. Satorius Executive Director for Operations FROM: Eric J. Leeds, Director /RA/

Office of Nuclear Reactor Regulation

SUBJECT:

STATEMENT OF VIEWS REGARDING APPEAL OF DIFFERING PROFESSIONAL OPINION CONCERNING DPO 2013-002 On July 19, 2013, in accordance with Management Directive 10.159, The NRC Differing Professional Opinions Program, a differing professional opinion (DPO) concerning seismic issues at the Diablo Canyon Nuclear Power Plant (DCNPP) (DPO-2013-002) was submitted.

On September 3, 2013, I established a DPO Ad Hoc Review Panel (the Panel) and tasked them to meet with the submitter, review the DPO submittal, and issue a DPO report, including conclusions and recommendations, to me regarding the disposition of the issues presented in the DPO.

On April 3, 2014, after reviewing the applicable documents, completing internal reviews of relevant individuals and completing their deliberations, the Panel issued their report to me. On May 29, 2014, I issued a closeout memorandum to the submitter documenting my decision regarding the DPO. On June 23, 2014, the submitter submitted an appeal to you regarding the DPO and my decision. This memorandum is to provide you with my views regarding statements in the appeal.

After reading the appeal, the submitter reiterated his stance on the reasons the DPO was originally submitted. I think its extremely important to note that the submitter continues to agree with the Panels conclusion that issues raised in the DPO did NOT result in a significant or immediate safety concern. The safety of the DCNPP is not in question. However, the submitter did not include any new, safety significant or other information that would cause me to alter my disposition of the DPO.

I also think it is important to note that the submitters DPO illustrates the need for the Agency to generically resolve how changes to external natural hazard parameters are processed by both licensees and the staff. This work is currently underway with regard to seismic and flooding hazards in response to the Fukushima accident. I expect the outcome of the staffs work on seismic and flooding issues will result in a well-defined process for both licensees and the staff to follow in the future, and this will help prevent a recurrence of the issues raised in the submitters DPO.

Please contact me if you have any questions regarding this Memorandum.

cc: M. Johnson, OEDO D. Dorman, NRR R. Pedersen, OE

ML14177A613 OFFICE NRR NAME ELeeds DATE 06/27/14 Document 7 - DPO Submitters Appeal Presentation

Diablo Canyon Seismic Issues Appeal of DPO 201302 Decision 1

DPO Issues:

The NRC failed to enforce:

• 10 CFR 50.59 requirement that Pacific Gas and Electric (PG&E) obtain an amendment to the Operating License prior to incorporating new seismic information into the FSARU.

• Plant Technical Specifications following inadequate demonstration of operability:

- New seismic information resulted in greater stress than plant structure, system and components (SSCs) were qualified.

- New seismic stresses exceeded the ASME Code acceptance limits for the reactor coolant pressure boundary (RCPB).

2

Diablo Canyon Seismic Design and Licensing Bases General Design Criteria (GDC) 2 & Part 100, Appendix A: Certain SSCs remain functional following the maximum earthquake potential considering the local geology and seismology:

• Integrity of the reactor coolant pressure boundary (ASME Code acceptance limits),

• Capability to shut down the reactor and maintain it in a safe shutdown condition, and

• Capability to prevent or mitigate the consequences of accidents (seismic qualification of plant SSCs).

3

Diablo Canyon Design and Licensing Bases Safe shutdown earthquake (SSE) design bases :

• Safety analysis (10 CFR 50.34) developed the Double Design Earthquake (DDE, 0.4 pga).

• Demonstrated that the GDC 2 functional requirements were satisfied for the maximum ground motion based on the local geology and seismology.

• Important to safety SSCs (RG 1.29, Seismic Design Classification) qualified to the DDE spectra.

• RCPB (Class 1 Systems) qualified to ASME,Section III, for DDE plus accident loads (10 CFR 50.55a).

4

Diablo Canyon Design and Licensing Bases Hosgri Fault discovered during plant construction:

• Licensee concluded that the fault was not capable per Part 100, Appendix A, and excluded the ground motion from the SSE safety analysis.

• PG&E prepared the Hosgri Evaluation (HE) in response to an NRC question during plant licensing.

• HE demonstrated that the plant could safety shutdown following 7.5 M on the Hosgri fault (0.75 pga) 5

Diablo Canyon Design and Licensing Bases Hosgri Evaluation (HE):

• Used different assumptions, methodology, load combinations, and acceptance limits than the DDE/SSE.

- Did not assume coincidental accident or fire.

- Explicitly excluded RG 1.29 (SSCs qualified for the SSE).

- Some Code limits exceeded (nonliner effects).

- Included ASME ,Section III, calculations for the RCPB, excluding LOCA loads (no accident).

• For many SSCs, including the RCPB, seismic qualification was more limited by the DDE/SSE (0.4 pga) rather than the HE (0.75 pga).

6

Diablo Canyon Licensing Bases Long Term Seismic Program

• License Condition - Reevaluate the local seismologic within 10 years.

• Completed in 1988.

• NRC concluded that PG&E satisfied the License Condition (1991):

- Did not alter the plant design bases.

- Seismic qualification basis will continue to be the DD and the DDE (OBE & SSE) design basis plus the HE, along with associated analytical methods, initial conditions, etc.

7

New Seismic Information Seismic Reevaluation submitted to the NRC (2011):

• Concluded three local faults were capable of generating significantly greater ground motion (0.7 pga) than used to establish the facility SSE.

• PG&E submitted License Amendment Request (LAR) 201105 to change the method of evaluation used for the facility SSE from the DDE to the HE.

• The NRC concluded that the HE did not meet NRC requirements for the SSE. At the NRCs request, PG&E withdrew the LAR (2012).

8

NRR Disposition NRR PM directed PG&E to add the Shoreline fault to the FSARU as a lessor case of the HE.

• The Hosgri ground motions were previously demonstrated to have reasonable assurance of safety.

• Deferred further evaluations pending Fukishima Recommendation 2.1.

• Did not address other faults that exceeded the SSE.

9

DPO Panel

Conclusions:

The new seismic information did not reveal a significant or immediate seismic safety concern

• However, the DPO did not assert that significant or immediate safety concern existed at Diablo Canyon.

• The DPO was written to draw attention and promote correct actions following the agency's failure to enforce existing regulatory and statutory requirements.

• Adequate protection (nuclear safety) is presumptively assured by compliance with NRC requirements.

10

DPO Panel

Conclusions:

The staff followed its processes for technical specification operability of plant equipment and 10 CFR 50.59 evaluations.

• However, the Panels conclusions were based on a different facility design and licensing bases than presented in the DPO and the FSARU.

• The Panel inappropriately considered the HE as a facility SSE.

• The Panel Report did not offer explanation for the deviation.

11

What Does the Regulations Require?

10 CFR 50.71(e) and 10 CFR 50.59:

• These statutory requirements required PG&E to evaluate the new information against the facility as described in the FSARU.

• Ensures fidelity is maintained between the functional GDC requirements, the methods used to demonstrate that the GDCs were met (10 CFR 50.34 safety analysis, as presented in the License Application, as amended), and the plant technical specifications.

12

What Does the Regulations Require?

10 CFR 50.71(e) required PG&E to updated the FSARU with the new seismic information:

• The new information was developed by PG&E.

• The new ground motions were greater than those used in the safety analysis demonstrating that GDC 2 was satisfied (design basis controlling parameter).

• The new ground motions were also greater than those used to demonstrate that ASME Code requirements were satisfied for the SSE per 10 CFR 50.55a.

13

10 CFR 50.71 (e) contain all the changes necessary to reflect information and analyses submitted to the Commission by the applicant or licensee or prepared by the applicant or licensee pursuant to Commission requirement shall include the effects of all changes made in the facility or procedures as described in the FSAR; all safety analyses and evaluations 14

10 CFR 50.59

• Change: A modification or addition to, or removal from, the facility or procedures that affects a design function, method of performing or controlling the function, or an evaluation that demonstrates that intended functions will be accomplished.

• Facility as described in the FSAR:

- Evaluations or methods of evaluation included in the FSAR for such SSCs which demonstrate that their intended function(s) will be accomplished.

- Submitted in accordance with §50.34, as amended and supplemented, and as updated per the requirements of Sec. 50.71(e).

15

10 CFR 50.59

• Departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analyses:

- Changing any of the elements of the method described in the FSAR unless the results of the analysis are conservative or essentially the same; or

- Changing from a method described in the FSAR to another method unless that method has been approved by NRC for the intended application.

16

What Does the Regulations Require?

10 CFR 50.59 required PG&E to obtain a an amendment to the Operating License for the FSARU:

• Modification of the design basis controlling parameter (ground motions) to the existing safety analysis resulted in exceeding acceptance limits for ASME Code and current SSC seismic qualification.

• This resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety, (NEI 9607).

17

What Does the Regulations Require?

10 CFR 50.59 also required PG&E to obtain an amendment to the Operating License to change the SSE methodology from the DDE to the HE:

• The HE was less conservative than the DDE.

• Results in a departure from a method of evaluation described in the FSAR used in establishing the design bases or in the safety analysis (NEI 9607).

• Considering the Shoreline fault a lessor case of the HE attached the HE method of evaluation to the SSE.

18

DPO Panel Concluded:

The new ground motions were bound by the Hosgri and the Long Term Seismic Program.

• The DPO did not dispute that the new ground motions were bound by the HE and the LTSP.

• However, this information was not relative since neither the HE nor the LTSP were part of the facility SSE (GDC 2) design bases.

• The DPO Panel Report did not include the based for using the alternant design and licensing bases.

19

The HE was not the Facility SSE The original and current FSARs were clear:

• The DDE, with supporting safety analysis, was used to demonstrate (GDC 2) SSE design basis.

• DDE/SSE design basis was implemented and controlled by the licensee using the facility QList.

• The DDE ground motions are used to establish SSC seismic qualification per PG&Es commitment to RG 1.29.

• The Hosgri was treated as a licensing basis commitment (beyond design bases), not attached to GDC 2 (or any other regulatory requirement).

20

The HE was not the Facility SSE

• Prior to the Panel Report, the DDE/SSE design basis was not in dispute.

• NRR and PGE initially agreed that an amendment to the license was required because the FSAR did not described the HE as the SSE.

• After the staff determined that the HE did not meet NRC requirements for the SSE (nonacceptance of LAR 1105), NRR implemented a work around to the license amendment process.

21

SSC Seismic Qualification was More Limiting for the DDE

• For a given ground motion, the DDE methodology will always produce greater SSC seismic stress that the HE.

• Ground motion alone does not established seismic design basis (NRC SSER 7). Equally important are other factors:

- Methods of analysis

- Shape of spectra

- Damping values used

- Load combinations

- Initial conditions

- Acceptance criteria, including allowable stress 22

Consequence of the Failure to Obtain an License Amendment The NRR PMs action subverted the required License Amendment Request public notice and hearing opportunities per 10 CFR 50.91:

• Substantial stake holder interest in Diablo Canyon seismic issues.

• Inadequate NRC review of plant SSC response to the higher ground motions.

• Adversely affects public perception of NRC as regulator.

• Established a new precedent for discovery of conditions outside of the existing design bases.

23

DPO Panel Concluded:

The staff followed its processes for technical specification operability of plant equipment.

• However, the Panel incorrectly assumed that the HE satisfied GDC 2 safety analysis (as described in the FSARU).

• As a result, the Panel Report did not address the specific technical issues raised in the DPO associated with use of the HE or LTSP as an alternative analytical method for determining operability.

24

What does the License Require?

Plant Technical Specifications required important to safety SSCs to be operable:

• Equipment needed to prevent or mitigate an accident must be capable of performing required safety functions following the SSE.

• The new seismic information called into question if this SSC functional requirement can be still be met at the higher ground motions.

• Applying the new ground motions to the existing SSE safety analysis resulted in stress exceeding the seismic qualification limits of important to safety SSCs.

25

What dose the License Require?

PG&E did not evaluate the new information against the SSE/DDE:

• Used the HE as an alternative analytical method.

• Not permitted per IMC 0326 (Appendix C.4):

- The HE methodology will always overpredict SSC performance when compared to the FSARU SSE methodology.

- For a given ground motion, the DDE/SSE will always be more limiting for seismic qualification.

- Successful demonstration of Technical Specification SSC operability is required for continued reactor operation.

26

DPO Panel Concluded:

The new information by itself did not alter the FSARU approach to maintain both he DDE and HE as failed conditions with respect to seismic component and piping analysis:

• The Panel concluded that either the HE or the DDE established the limiting loads for ASME acceptance.

• Since the new ground motions were less than those assumed in the HE, then all ASME Code requirements were satisfied.

27

What Does the Regulations Require?

10 CFR 50.55a required PG&E to met ASME,Section III, Code requirements for the RCPB (Class 1 Systems):

• SSE plus accident loads must be less than acceptance limits (Service Level D).

• The new seismic information resulted in a greater maximum (credible) earthquake potential than described in the FSARU SSE safety analysis. This rendered the 50.34 safety analysis nonconforming with the requirements of GDC 2, (Criteria III Design Control, and XVI Corrective Actions).

• The nonconforming supporting safety analysis was used as input for satisfying 10 CFR 50.55a.

28

What Does the Regulations Require?

10 CFR 50.55a required PG&E to meet ASME,Section III, Code requirements:

• Applying the new ground motions (design bases controlling parameter) to the existing SSE safety analysis resulted in exceeding Code acceptance limits.

• The Code did not include provision for substitution of the HE for the SSE for seismic inputs.

• Meeting ASME Code acceptance limits for the RCPB (Class 1 Systems) was required for continued reactor operation (IMC 0326, Appendix C.11).

29

DPO Panel Concluded:

The lack of a formal regulatory guidance for new information appeared to contribute creating differing interpretations for the potential significance.

• The DPO limited the potential significance of the new information to the nexus between compliance and safety.

• The Enforcement Manual and Significance Determination Process should have been used to establish the actual safety significance of the issues and ensure adequate corrective actions.

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The Current Regulatory Framework Ensures Continuity Between:

• GDC functional requirements and the design bases.

• FSAR safety analysis (10 CFR 50.34) demonstrating that the design bases satisfies the GDC functional requirements.

• Part 50, Appendix B, ensures that this design bases is maintained by the facility safety analysis and design control for individual plant SSCs.

• 10 CFR 50.71(e) ensures new information that affects the design bases/safety analysis is updated in the FSARU.

• 10 CFR 50.59 ensures fidelity is maintained between the design bases and FSAR safety analysis methods (GDCs).

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Additional Formal Regulatory Guidance Was Available Supplemental information reinforced agency regulations to address nonconforming conditions associated new information related to natural phenomena or the failure to meet a GDC as described in the current licensing bases (FSARU):

• Letter (Leeds), supplemental information related Recommendation 2.1

• Regulatory Issue Summary RIS 201305,

• IMC 0326, Appendix C1,

• Past enforcement actions (Watts Bar Flooding) 32

DPO and Nonoccurrence Processes were Ineffective Nonconcurrence NCP 201201 addressed PG&Es inadequate operability evaluation. The Agency did not respond to the technical issues:

• Code compliance,

• Inappropriate use of HE as alternate analytical method.

The DPO Panel created a new facility design and license bases. The Agency did adequately address the issues raised in the DPO:

• Specific criteria in 10 CFR 50.59/NEI 96.07 33

Summary DPO Panel did not fully consider the original and current facility design and licensing bases (FSAR):

• Led to incorrect conclusions related to compliance with 10 CFR 50.71(e) and 10 CFR 50.59.

• FSARU ambiguities require corrective action and do not provide an adequate bases for deferring enforcement action.

• The lack of an immediate or significant safety issue does not provide adequate justification for failing to enforce statutory and license requirements.

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Summary

• The NRR PMs bypassed the license amendment process:

- A reasonable assurance of safety, was inconsistent with current regulatory requirements (10 CFR 50.59).

• Agency work around of the licensee amendment process created potential for safety significance:

- Inadequate agency review of the impact of new seismic information on plant SSCs (10CFR 50.59)

- Subverting the required notice and hearing opportunity (10 CFR 50.91) 35

Summary

• Continued failure to enforce plant technical specification requirements.

• Improvements are needed to enhance agency accountability for the nonconcurrence and DPO process decisions.

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Recommend Actions

1. Disapprove the DPO 2013002 Panel Report decision:

- The Panel Report applied an incorrect design and licensing bases when addressing the DPO compliance issues.

- Use of the correct facility design and licensing based substantiates the issues raised in the DPO.

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

- The facility continues to operate outside the bounds of the current safety analysis and design bases.

- PG&E has not adequately demonstrated that all technical specification required SSCs are operable.

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Recommend Actions

3. Initiate a review to determine why the nonconcurrence and the DPO processes were not effective:

- Region IV response to NCP 201201 did not address the technical issues raised.

- DPO Panel created a new licensing bases to justify past inappropriate agency actions.

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Document 8 - DPO Appeal Decision