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Diablo Canyon Independent Safety Committee’s Evaluation of Pressurized Thermal Shock and Seismic Interactions for a 20-Year License Extension at the Diablo Canyon Nuclear Power Plant – 15 February 2011

Concurred in by the three members of the DCISC at the DCISC Public Meeting in San Luis Obispo on 15 February 2011
Robert J. Budnitz
Peter Lam
Per F. Peterson

BACKGROUND: THE REQUEST FROM THE CEC

The request to the DCISC is on page 236 of the California Energy Commission report whose citation is “2009 Integrated Energy Policy Report”, Report CEC-100-2009-003-CTF (December 2009)

The specific request on page 236 is as follows:
“The Diablo Canyon Independent Safety Committee should evaluate reactor pressure vessel integrity at Diablo Canyon over a 20‐year license extension and recommend mitigation plans, if needed. This review should consider the reactor vessel surveillance reports for Diablo Canyon in the context of any changes to the predicted seismic hazard at the site.”

TECHNICAL BACKGROUND

Pressurized water reactors (PWRs), like the two units at the Diablo Canyon Power Plant (DCPP), operate at high pressures and temperatures (approximately 2200 psi and 600 degrees F). The reactor vessel (RV), which contains the nuclear reactor core, is fabricated out of steel plates, around 8 inches thick, which are formed into curved and dished sections and then welded together. The RV has a number of penetrations, including for control-rod drives, instrumentation lines, and four hot-leg and four cold-leg nozzles that connect to the steam generators and primary pumps. These nozzles and their piping provide circulating water to cool the reactor core. The emergency core cooling system (ECCS) is designed to provide sufficient coolant injection to prevent fuel damage in the event that any of these connecting lines or pipes were to break, including a large break of either a hot-leg or cold-leg pipe. The possibility that the RV itself might break is considered to be so unlikely as to be a beyond-design basis event, for which the ECCS need not be designed. To assure that this reasoning remains correct, mechanisms that might cause failure of an RV have been investigated extensively, particularly those that might result in brittle failure of the RV or its welds due to the accumulated effects of neutron irradiation.

The metallurgy of the reactor vessel of a PWR and of its welds is selected so that at operating temperatures the vessel and its welds are ductile. At relatively cool temperatures the vessel (or a weld) becomes brittle. The temperature at which this transition occurs, in a narrow range of only a few degrees, is known as the “Reference Temperature for nil-ductility transition” (called RTNDT). A brittle vessel or weld represents a threat to safety, because under sufficiently high stresses such a vessel or weld could lose its integrity and fail, resulting in potentially severe damage to fuel and release of radioactive material into the reactor containment. One event with the potential to generate sufficiently high stresses involves the injection of cold ECCS water under pressurized conditions, causing high, localized thermal stresses in the RV. Because no large vessel is entirely free of minor cracks and other flaws, the phenomenon of concern would be that under high stresses one of these existing, small cracks might grow to become a major flaw or a breach of the vessel’s integrity.

To assure that such failures do not occur, the NRC has established requirements that govern the fabrication of these vessels and welds, their maintenance, and their periodic inspection. As mentioned, it is not possible to fabricate any such vessel or welds without some small cracks and other flaws. There are industry and NRC requirements that assure that no cracks or flaws are larger than the size range that is known to represent a threat to the integrity of the overall vessel. During a plant’s operating life various periodic inspections and analyses are made to maintain adequate assurance about this aspect of safety. All nuclear plants maintain records of the small cracks and flaws identified after fabrication, the periodic measurements of them, whether any new ones are identified, and whether those being tracked remain well within tolerances permitted for safe operation.

In addition, every vessel contains surveillance specimens, pieces of metal that are placed inside the vessel, made of the same material as the vessel, and subjected to the same or a similar thermal and radiation environment as the vessel itself. These specimens (called coupons) are extracted periodically for examination in a laboratory setting using advanced techniques that can detect changes in the material that would be difficult to detect by in situ measurements in the vessel itself. There is a concern that at some nuclear plants an extension of operation for another 20 years will mean that there will not be enough specimens to continue this surveillance effectively. The NRC is currently examining this issue. At the present time this is not an issue for the Diablo Canyon Power Plant (DCPP) vessels. The DCISC has learned from DCPP plant staff that the DCPP plant possesses enough metallic coupons, either in the reactor itself or removed and now stored in the spent-fuel pool, to support the plant’s need to understand potential radiation damage to the reactor vessels out for the full 60-year proposed lifetime of the plant if NRC grants a license extension. Specifically, the irradiation experience from the coupons already in-hand at DCPP extends in some cases to the equivalent neutron fluence for about 55 EFPY (effective full power years), close to what would be needed for a 60-year operating lifetime. The coupons with the highest neutron fluence exposures have received 55 EFPY (even though the plant itself has run for less than 25 years) by having been placed in a higher neutron flux field inside the reactor core than the fluence that the vessel walls have experienced. If the exposures on these coupons are accepted by the NRC as a valid basis, then the DCPP plant already has almost enough irradiation experience with the coupons in-hand to support their need out to the proposed 60 years.

NRC regulations and industry consensus codes also require that the vessel and welds be designed to withstand a wide range of off-normal operating conditions that might occur, including conditions whose occurrence would be quite infrequent. One of these is additional stresses induced by the occurrence of a large earthquake. Analysis has been done and measurements made to assure that such additional earthquake-induced stresses are sufficiently small, compared to the stresses imposed by a postulated pressurized thermal shock (discussed below), that they remain within elastic limits and thus are not a threat to reactor vessel integrity, and in fact they are not. Industry analyses and NRC reviews all agree that there is significant margin in all operating vessels for these possible events. This includes the two vessels at DCPP.

While the stresses induced in the RV under normal operation and during earthquakes remain within elastic limits, a phenomenon referred to as “thermal shock” can generate substantially higher localized stresses that can potentially exceed elastic limits. Thermal shock may occur if cold water is injected into a hot and highly-pressurized vessel in response to one of several possible upset conditions. For example, under certain off-normal conditions following a pipe break, or a stuck open safety or relief valve, or other so-called loss-of-coolant-accident (LOCA) events, it is necessary to inject water into the vessel to assure that the core remains covered and cooled. This injection water is typically much cooler than the operating temperature of the vessel, and this cooler water can both induce thermal shock directly in the hotter highly-pressurized vessel and also it can cool the vessel locally, creating large local thermal stresses and making the vessel itself less ductile, or in a worst case potentially forcing the material into a transition to a brittle state by cooling it below the RTNDT transition temperature.

Such a transient, known as a “pressurized thermal shock” or PTS, is an accident sequence that the NRC and the industry have studied extensively over the years to assure that it does not happen. The threat is important for PWRs (pressurized water reactors) like Diablo Canyon but it is essentially absent from BWRs (boiling water reactors), because the latter operate at significantly lower pressures where the stresses leading to the shock phenomenon do not occur.

The reactor vessel and weld materials are, of course, subjected to radiation, especially neutron irradiation, during the reactor’s operation. Under neutron irradiation, the RTNDT temperature increases slowly over the years. With this increase, there is less “margin” against the transition from a ductile to a brittle state than in an unirradiated vessel, because the metal needs to be cooled less than in a new vessel to reach the RTNDT transition temperature. This phenomenon, which has been known for decades, has been the subject of significant research study over the years. One of the NRC’s requirements is that the gradual change in material properties due to neutron irradiation be examined periodically to assure that a safe operating condition continues to exist.

Early studies indicated that it would take many years of operating service for this phenomenon to become a threat, but a few cold-water-injection events in operating reactors in the late 1970s led the NRC to intensify both its research into PTS phenomena and its regulatory attention to PTS. This ultimately led to a special NRC regulation in 1983 addressing PTS, 10 CFR Part 50.61, “Fracture toughness requirements for protection against pressurized thermal shock events” (Ref. 1). Based on the knowledge of the time, this regulation required the operators of the PWRs to perform certain analyses, make certain measurements, and report regularly to the NRC about their findings. NRC published regulatory guidance that provided additional information for the power plants about how to deal with this issue (Ref. 2). However, the knowledge about PTS phenomena and also about other features of the reactors was not as advanced at that time as it is now, so the regulation made certain conservative assumptions just to remain with high assurance “on the safe side”. This was a prudent approach that the entire engineering community endorsed, and PWRs around the world have precluded PTS scenarios ever since by following the regulatory prescriptions. There was also very high interest abroad (Ref. 3), and there continues to be extensive work on the topic.

ADVENT OF NEW TECHNICAL KNOWLEDGE, FOLLOWED BY A NEW NRC REGULATION

During the 1990s, additional research provided new knowledge about PTS and the potential accidents that might lead to it. This research was in three different areas: (a) Better understanding was developed about the specific types of off-normal scenarios that might lead to the PTS phenomenon, specifically the likelihoods of the various scenarios. (b) Better understanding was developed about the thermal and thermalhydraulic conditions that would occur during the various off-normal scenarios, and (c) Better understanding was developed about the metallurgical properties of the vessels and welds and of their responses to PTS-type events.

All three of these lines of evidence came together in the late 1990s and a consensus emerged in the technical community that the existing NRC regulation and its associated regulatory guidance were substantially conservative. Specifically, the NRC required certain analyses and measurements that later knowledge revealed were not necessarily required, and in some circumstances the regulation might even have required some reactors to make certain operational changes in how they could operate their reactors that were not needed to assure safety. For example, one way to gain additional “margin” against PTS is to change the way fuel is loaded into the reactor. The change would modify how the fuel is placed in the reactor and how it is irradiated, so as to decrease the neutron fluences striking the vessel walls. Such changes, which are otherwise undesirable because they change the power distribution in the reactor core and can decrease reactor thermal efficiency, would be required in some reactors late in their operating life to meet the existing PTS rule but perhaps would not actually be required to maintain adequate safety.

To assure that their regulations and guidance reflected the latest science, the NRC in the late 1990s began a research program to investigate all three of the technical issues cited above, covering the likelihood and characteristics of the PTS scenarios, the understanding of thermal and themalhydraulic aspects, and the understanding of the metallurgical aspects. These NRC research programs, complemented in some areas by industry and academic research both in the US and abroad, led the NRC to propose a revised PTS regulation, which was ultimately adopted in early 2010, after several years of review by experts in all of the various technical fields. This new rule, known as 10 CFR 50.61a (Ref. 4), is now in force, but it is optional, meaning that any operating reactor owner can either continue to abide by the former regulation 50.61 or choose to use the more modern regulation 50.61a. (One reason why an owner might not choose to use the new approach is that it imposes certain procedures and costs that an owner might not think are necessary – the overwhelming opinion in the technical community is that the long-standing rule 50.61 maintains adequate safety for all reactors, and most will not need to invoke the new regulation, although some may wish to rather than make the operational changes.) Indeed, the revised NRC regulatory approach in 50.61a has been met with wide technical approval.

An extensive body of technical literature was developed by the NRC in the form of research articles and regulatory reports to support their new regulation (Ref. 5, 6, 7, 8).

One characteristic of the new NRC regulation is that it uses a more modern regulatory philosophy known as “risk informed regulation”, with explicit risk-type targets as part of the regulatory framework in some instances. However, the NRC explicitly stated that the new approach was intended to be neither more stringent nor less stringent in terms of the overall safety level that a reactor would achieve by meeting either of the two PTS rules. That is, the new rule was not intended as either a tightening or a relaxation of the overall safety level required of the plants that would opt to meet the new rule instead of continuing to meet the old rule.

LICENSE RENEWAL ISSUES

All of the original NRC licenses for the large power reactors were issued for 40 years. However, NRC’s regulations allow a reactor to obtain a 20-year extension to its operating license if it meets certain regulatory requirements, mostly involving aging of structures and components. Among those requirements is that a new analysis must be done to assure that PTS phenomena do not threaten reactor safety over the 60 year period. Because additional years of operation mean significant additional neutron irradiation of the vessel and the weld materials, an analysis is needed to assure that, even at the end of the 60th year of proposed operation, the PTS threat will remain within regulatory limits.

As mentioned, to support the adoption of the new NRC regulation, the NRC Office of Research sponsored a series of research studies on all three of the technical aspects mentioned above, covering the likelihood and characteristics of the PTS scenarios, the understanding of thermal and thermalhydraulic aspects, and the understanding of the metallurgical aspects.

A large number of different potential accident sequences were studied by the NRC and its contractors, and included both accident sequences initiated by so-called internal plant faults and human errors, and accident sequences initiated by earthquakes.

PTS transient sequences initiated by so-called internal plant faults and human errors: The types of sequences that were found to pose the greatest potential PTS threat in the operating PWRs are transients involving primary-side faults that could cause a through-wall crack in the vessel (Ref. 5). The sequences of greatest concern were found to involve medium-diameter and large-diameter primary side pipe breaks, and stuck-open primary side valves that later reclose. For the pipe breaks, the fast cooling rates in some scenarios, combined with relatively low temperatures in the reactor downcomer region (arising from rapid depressurization and emergency injection of low-temperature makeup water directly into the primary), combine to produce a possible high-severity transient event. For the scenarios involving a stuck-open valve that later recloses, the repressurization associated with the reclosure combined with low temperatures in the primary side can combine to produce a potential PTS scenario.

Transients involving the secondary side, such as main steam line breaks and stuck open secondary valves, were found not to produce a significant PTS threat, largely because there is a very low probability that these transients will produce sufficiently low temperatures on the primary side.

Each of the major potential scenarios was studied in detail (Ref. 5). This involved both thermalhydraulic analyses and metallurgical analyses. The overall conclusion was that the threat of any one of the PTS scenarios is small if certain metallurgical parameters are maintained in the vessel and its welds. Therefore, the new NRC rule (like the older one) contains provisions to assure that the metallurgical compositions of the materials remain in the safe region, along with other administrative provisions. The provisions in the new regulation, if followed, achieve the same very high level of assurance, but the advent of modern scientific knowledge has made the new regulation less onerous to meet.

PTS transient sequences initiated by earthquakes: The NRC also supported a separate study (Ref. 8) of potential accident sequences initiated by earthquakes (and other external hazards.)

While initially there was concern that seismic loads themselves could damage the reactor vessel, this has been studied carefully and the conclusion from these studies is that this is not an important safety issue, because the stresses induced in the reactor vessel even by large seismic loads fall well within the vessel’s strength regime, if the vessel is operated to follow the restrictions in the NRC’s regulations, including among others its PTS regulations. The vessel at its operating temperature will respond elastically to such loads, and failure from them is not a concern. This conclusion is concurred in by the broad engineering community, and is valid even in the hypothetical presence of potential seismic motions at DCPP much larger than could arise from earthquakes currently accounted for by NRC’s regulations.

Rather, the major concern is that an earthquake could initiate a cold-water safety injection leading to PTS. The NRC studied a few PWR reactors specifically, and then used these few plant-specific studies as benchmarks to determine the seismically-induced PTS risk for all of the PWRs now operating (Ref. 7, 8). Among the plants chosen for the external-hazard studies was Diablo Canyon, in part because the earthquake hazards at the Diablo Canyon site are higher than at any other US plant (along with the earthquake hazards at San Onofre, the other operating nuclear station in the high seismic environment of coastal California.) The NRC-sponsored studies concluded that, even though the seismic hazard at Diablo Canyon is as high as it is, and the earthquake-initiated scenarios examined are possible, the frequency of seismically induced PTS transients would be low compared to the frequency of other initiating events that could produce PTS, and the character of these earthquake-initiated PTS events would be similar to the character of PTS events initiated by other causes. This leads to the conclusion that the additional risk arising from potential seismically-induced PTS scenarios is not an important contributor to the overall PTS risk. Specifically, the conclusion was that the overall PTS risk including that arising from potential earthquake scenarios remains within NRC regulatory requirements and the vessel’s design basis, if the vessel and welds are managed according to NRC regulations.

ANALYSIS PERFORMED BY DCPP

As a major part of its application for a 20-year license extension, DCPP developed an extensive technical submittal (Ref. 9). One part of that submittal explicitly addresses the PTS issue. The DCPP submittal is backed up by other engineering analyses (Ref. 10 – 14), which include surveillance data and analyses as well as other information.

The DCPP analysis (Ref. 9) concludes that both Units can operate out to 60 calendar years and remain in compliance with the NRC’s PTS regulations. Because Unit 2’s vessel has different metallurgical properties than that of Unit 1, it has more margin than Unit 1, but according to DCPP’s analyses there is adequate margin for both units. DCPP believes that both units can meet the more conservative NRC regulation 50.61 out to 60 calendar years without invoking the additional benefit from using the newer 50.61a approach. However, PG&E agrees that for Unit 1 they are not completely sure that it can comply with 50.61 all the way to 60 calendar years, and the situation won’t be clarified until several years in the future. Therefore, the DCPP license application commits to careful experimental and analytical studies over the ensuing future years to assure this, and commits either to invoking 50.61a at sometime in the future if needed, or making other operational changes to assure compliance. An analysis subsequent to the license-renewal application, performed by DCPP staff, and supported in part by engineering studies performed by contractors (Ref. 10-13), concludes that using the 50.61a regulation, Unit 1 can comply fully out to the 60 calendar-year operating period. According to DCPP, Unit 2 has an even stronger safety case, that is, even more margin.

NRC REVIEW OF DCPP LICENSE RENEWAL

The NRC has recently completed its review of the DCPP application for a 20-year license renewal, and has issued its first-round “Safety Evaluation Report” (Ref. 15). This is the first round because the NRC staff has identified “eight open items and two confirmatory items that must be resolved” before the final SER is issued. However, none of these open items involves issues related to the technical topics here, each of which is considered resolved by the NRC staff.

Basically, the NRC staff concluded that the information furnished by PG&E about the pressure vessel and the program of surveillance and possible future intervention is adequate to support the license renewal. This NRC conclusion supports our own DCISC technical evaluation

THE NEW SEISMIC HAZARD INFORMATION – THE NEW “SHORELINE FAULT”

Recently, in 2008, studies in the vicinity of DCPP uncovered a new seismic feature, called the Shoreline Fault, that was not known before. Since its discovery, this feature has been the subject of intense study by PG&E, the NRC, the State of California, the US Geological Survey, and the seismic research community generally. The DCISC has followed developments as they have emerged, reviewed information developed by all of the various research teams, heard from the DCPP experts in our Public Meeting format as well as in separate technical review meetings, reviewed an interim 2009 NRC report (Ref. 16), and attended both an NRC public seismic workshop on 8-9 September 2010 and a PG&E Public Meeting on 19 January 2011, both in San Luis Obispo. We have also received the latest PG&E technical report on the subject of the new Shoreline Fault (Ref. 17), and have begun a review of it.

The concern is that perhaps this new seismic zone could produce earthquake ground motion at the site for which the DCPP plant was not designed. This concern is shared by everyone – by the plant, by the NRC, and of course by our Committee. If it is valid, the plant might require engineering backfits or even potentially might not be safe enough to continue operating.

The NRC has strict regulations that govern how a nuclear power plant must be designed to withstand earthquakes, and the design itself must be, by NRC regulation, adequate to cope with the seismic ground motion that is characteristic of the specific site. The regulations require consideration of all seismic source features, near and far, that might affect the power plant, and the approach in the regulations then uses the most likely one or more of these features as the basis for developing the requirements for the design. How the seismic energy propagates from source to site is also taken into account in the NRC regulations, on a site-specific basis. The design regulations have several specific technical requirements governing not only the buildings and the equipment but also the foundation and other engineered and natural characteristics of the site, including in the subsurface below the structures. The specific set of design requirements, as prescribed by NRC’s regulations, are known as the “seismic design basis” for that nuclear power plant.

DCPP’s studies to date, as reported to the DCISC at our public meetings and to the public at NRC’s and PG&E’s public meetings, have concluded that the additional seismicity represented by the new Shoreline Fault feature does not require changes to the plant’s NRC-mandated seismic design basis. Specifically, DCPP’s experts conclude that the current DCPP design basis against earthquake hazards, which is dominated by the hazard from the Hosgri Fault zone, remains an adequate design basis. They conclude that if the seismic design basis were developed anew using NRC’s prescriptive regulatory approach and accounting for all new information (including both the new Shoreline Fault feature and other new information such as an improved understanding of ground-motion propagation in the surrounding geology), the new design basis would not require any changes in the current plant as it sits. In 2009, the NRC staff evaluated the interim information available at that time, and agreed with the DCPP conclusion, although it stated that this was only an interim NRC position (Ref.16).

Crucially, DCPP’s staff has recently performed new analyses (Ref. 17) to assess the impact on the plant design basis from seismic activity on the Shoreline Fault feature by itself, and separately from a scenario in which both the Hosgri fault and the Shoreline fault are assumed to produce seismic energy together, whether caused by one fault acting on the other, or caused by some external seismic or tectonic force that activates both faults together. DCPP’s preliminary results indicate that, because of the geometry of the two fault zones, the potential for an earthquake on the Hosgri fault to produce a splay rupture onto the Shoreline fault is very unlikely. According to these results, if this joint activity were considered as the so-called deterministic earthquake event for use in the NRC regulations as the source of the design basis for the site, its likelihood is too low to be used in that prescribed NRC regulatory process. The latest PG&E analysis (Ref. 17) shows that the Shoreline Fault adds a modest amount to the ground motion, but that overall the seismic ground motion is smaller than it was thought to be decades ago when the original analysis was done and the design was approved and built. This is principally because for a given earthquake postulated to occur nearby, the attenuation of the ground motion as it travels from the source to the DCPP site is found by PG&E to be greater than previously thought. PG&E concludes that the decrease in ground-motion due to the better attenuation information is larger than the modest increase arising from the new Shoreline Fault. The net effect is a decrease, which according to PG&E translates into an overall smaller risk of core damage from earthquakes, as shown by a new PG&E SPRA (seismic probabilistic risk analysis). PG&E’s recent report supports this finding with technical data and analysis that uses the latest site information and up-to-date understanding of ground-motion attenuation.

The NRC staff has not yet evaluated this new DCPP analysis, and neither has the DCISC. Even though these results are preliminary, they meet an essential need to examine the impact of the Shoreline Fault by itself and also the joint impacts of the Hosgri and Shoreline faults together.

These DCPP staff efforts provide additional information to help determine whether the recently discovered Shoreline Fault zone poses a threat exceeding NRC’s regulatory limits. The DCPP team has developed a plan for further seismic studies, has submitted it to the NRC for review, and is continuing to pursue those additional studies. The NRC staff is following these studies closely, and has stated that it will continue to re-evaluate its interim position as the additional information comes in.

DCISC EVALUATION OF THE ABOVE

The Diablo Canyon Independent Safety Committee has been following the technical issues associated with the new Shoreline Fault feature, with the application by PG&E for a license renewal, and with the specific issue of PTS as it is affected by both the new seismic information and the proposed license extension. Our work has included studying a variety of technical reports, from both industry and governmental sources as well as foreign studies; meeting with DCPP personnel during our regular monthly Fact Finding trips to the plant; hearing and reviewing presentations from DCPP during our periodic public meetings; discussing various technical issues with the relevant NRC experts and their contractors; reviewing NRC regulations, industry codes and standards, and standard industry practices; and relying on our experience and expertise to pull the above together into what we believe is a coherent picture of the current situation. The DCISC believes that although our conclusions are always “preliminary” in the sense that they are open to revision based on new technical information, there is enough information available now to support some important conclusions. And in particular, in recognition that active technical work is underway to develop more information about the newly discovered Shoreline Fault feature, we will be alert going forward to whether any of that new information will require us to reevaluate the conclusions reached as of now.

With the above as background, the DCISC has reached the following conclusions, based on its review of the information available to it. As mentioned, this includes not only information provided by PG&E in both its license renewal application and in supplemental studies made available to the DCISC, but also information available from the nuclear industry, and the NRC and its contractors. The DCISC has performed its own evaluation of all of this information, and has reached several conclusions. All of these are interim conclusions, in the sense that the DCISC will continue to monitor all new technical developments as they occur. Specifically, we have concluded as follows.

DCISC CONCLUSIONS

Based on our review, we do not believe that any special mitigation measures are needed at this time, beyond those that are already underway or required, to assure ongoing safety of the plant vis-à-vis these joined issues of the seismic design, the design and operational regime to cope with PTS, and their nexus. The NRC has placed certain requirements on the DCPP plant in the normal course of their regulatory oversight, will be requiring more of the same and some new and different requirements if a 20-year license renewal were to be granted, and the plant itself has ongoing programs that follow industry codes and standard practices. These are currently adequate, in our opinion. The DCISC will of course continue to follow events as they unfold, including both technical information and NRC regulatory positions as they might emerge or change.

DCISC FOLLOW-ON ACTIVITIES

As mentioned, these are all interim DCISC conclusions, in the sense that as new information is developed any of them is subject to renewed evaluation. In particular, as a follow-up to the work done so far on this set of issues, we will undertake the following:

All of the above would be a part of our normal DCISC scope to review operational safety at DCPP, but because of the special inquiry made by the California Energy Commission, we will be especially alert about these issues.

REFERENCES

  1. NRC Regulation, Title 10 Code of Federal Regulations Part 61 (10 CFR 50.61), “Fracture toughness requirements for protection against pressurized thermal shock events” (1983)
  2. NRC Regulatory Guide 1.154, “Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis” (1987)
  3. OECD Nuclear Energy Agency (Paris, France), “Regulatory Practices on Pressurized Thermal Shock”, Report NEA/CSNI/R(91)7 (1991)
  4. NRC Regulation, Title 10 Code of Federal Regulations Part 61a (10 CFR 50.61a), “Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events” (2010)
  5. NRC Report NUREG 1806, “Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61)” in 2 volumes (2007)
  6. NRC Report NUREG 1874, “Recommended Screening Limits for Pressurized Thermal Shock” (2007)
  7. D. Whitehead et.al., Sandia Letter Report to the NRC, “Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants” (2004)
  8. A. M. Kolaczkowski, D. Kelly, and D. W. Whitehead, Sandia Letter Report to the NRC, “Estimate of External Events Contribution to Pressurized Thermal Shock (PTS) Risk” (2004)
  9. DCPP, “License Renewal Application, Diablo Canyon Power Plant Unit 1 and Unit 2, Facility Operating License Nos. DPR-80 and DPR-82” (2009)
  10. Westinghouse Electric Company, “Analysis of Capsule V from PG&E Diablo Canyon Unit 1 Reactor Vessel RadiationSurveillance Program”, Report WCAP #15958 (2003)
  11. Westinghouse Electric Company, “Analysis of Capsule V from PG&E Diablo Canyon Unit 2 Reactor Vessel Radiation Surveillance Program”, Report WCAP #15243 (2000)
  12. Westinghouse Electric Company, “Fast Neutron Fluence and Neutron Dosimetry Evaluations for the Diablo Canyon Unit Two Reactor Pressure Vessel”, Report WCAP #15782 (2001)
  13. Diablo Canyon Power Plant Design Calculation 9000007298 R1, “Reactor Vessel Fluence Evaluation in Support of License Renewal Application” (2009)
  14. Diablo Canyon Power Plant Design Calculation 9000040901 R0, “Reactor Vessel PTS/USE (upper shelf energy) Calculation at 54 EFPY Using New Pressurized Thermal Shock(PTS) Rule in Support of License Renewal Application” (2009)
  15. U.S. Nuclear Regulatory Commission, “Safety Evaluation Report With Open Items Related to the License Renewal of Diablo Canyon Nuclear Power Plant, Units 1 and 2, Docket Nos. 50-275 and 50-323, Pacific Gas and Electric Company” (January 2011)
  16. NRC Research Information Letter 09-001, “Preliminary Deterministic Analysis of Seismic Hazard at Diablo Canyon Nuclear Power Plant from Newly Identified ‘Shoreline Fault’” (2009)
  17. Pacific Gas and Electric Company, Diablo Canyon Power Plant, “Report on the Analysis of the Shoreline Fault Zone, Central Coastal California”, Report to the U.S. Nuclear Regulatory Commission, (January 2011)

For more information about DCISC contact:

Diablo Canyon Independent Safety Committee
Office of the Legal Counsel
857 Cass Street, Suite D, Monterey, California 93940
Telephone: in Califonia call 800-439-4688; outside of California call 831-647-1044
Send E-mail to:dcsafety@dcisc.org