Review of SONGS License Amendment Request by COPS; Petition to Intervene and Request for a Hearing

Citizens Oversight (2012-08-29) Ray Lutz

This Page: https://copswiki.org/Common/M1295
More Info: San Onofre License Amendment Intervention, Shut San Onofre

See main subproject here: San Onofre License Amendment Intervention

Latest News

A petition to intervene and request a hearing has been submitted as of Oct 17, 2012 by COPS. Read the full petition here:

• Petition to Intervene by COPS
• News Release on Petition to Intervene
• 2012-10-19_NRC_Memo_referring_request.pdf: NRC Memo which confirms reception and refers the request to Administrative Judge E. Roy Hawkins

People who are on the contact list for this submission

Last Name	First Name	Email	Affiliation	Web Page	Info
Bessette	Paul M.	pbessette@morganlewis.com	Morgan Lewis Company	Bio	Mr. Bessette represents nuclear utility clients in a variety of licensing, regulatory, adjudicatory, and litigation matters.
Burdick	Stephen	sburdick@morganlewis.com	Morgain Lewis Company	Bio	Mr. Burdick represents electric utilities and other clients in the nuclear industry on a variety of regulatory and litigation matters before the Nuclear Regulatory Commission (NRC), other state and federal agencies, and the federal courts.
Frantz	Steven P.	sfrantz@morganlewis.com	Morgan Lewis Company	Bio	Mr. Frantz represents and counsels electric utilities, manufacturers of reactors and materials licensees on the regulation and licensing of nuclear power plants, as well as other facilities regulated by the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE).
Freeze	Mary	mfreeze@morganlewis.com	Morgan Lewis Company	N/A	N/A
Hawkens	Roy	roy.hawkens@nrc.org	NRC	N/A	N/A
Hearing Docket		hearingdocket@nrc.org	NRC	N/A	N/A
Kanatas	Catherine	catherine.kanatas@nrc.org	NRC	N/A	N/A
Martin	Circe	ogcmailcenter@nrc.org	NRC	N/A	N/A
OCAAMAIL		OCAAMAIL@nrc.org	NRC	N/A	N/A
Roth	David	david.roth@nrc.org	NRC	N/A	N/A
Smith	Maxwell	maxwell.smith@nrc.org	NRC	N/A	N/A
Sutton	Kathryn M.	ksutton@morganlewis.com	Morgan Lewis Company	Bio	Kathryn M. Sutton is a partner in and the practice group leader of Morgan Lewis's Energy Practice. Ms. Sutton represents nuclear utility clients in licensing, regulatory and adjudicatory matters.
Walker	Antoinette	awalker@morganlewis.com	Morgan Lewis Company	N/A	N/A
Weisman	Robert	robert.weisman@nrc.org	NRC	N/A	N/A
Williamson	Edward	edward.williamson@nrc.org	NRC	N/A	N/A

Summary

The licensee ( Southern California Edison Company) submitted a license amendment request (LAR) for SONGS, Units 2 and 3, dated July 29, 2011, requesting approval to convert the Current Technical Specifications (CTS) to be consistent with the most recently approved version of the Standard Technical Specifications (STS) for Combustion Engineering Plants, NUREG-1432.

This review was conducted by Citizens' Oversight Projects; Reviewing Engineer was Ray Lutz.

After review, it is apparent that the primary change to the documents is to adopt a new "Risk-Informed Method for Control of Surveillance Frequencies" instead of the fixed frequencies specified in the Technical Specifications. The Surveillance Test Intervals (STIs) are then tracked by operator-controlled documents using a procedure designated by the NRC:

• "Risk-Informed Method for Control of Surveillance Frequencies" http://pbadupws.nrc.gov/docs/ML0713/ML071360456.pdf

Thus moving a spec to another document does not change it right away, but over time, the process of the SFCP (Surveillance Frequency Control Program) will be to gradually reduce surveillance rates. There does not seem to be a method to increase surveillance rates above the original rate, even if there is an unusual event tied to the system of interest.

B. Submitting Comments

Please include Docket ID NRC-2012-0192 in the subject line of your comment submission, in order to ensure that the NRC is able to make your comment submission available to the public in this docket.

The NRC cautions you not to include identifying or contact information in comment submissions that you do not want to be publicly disclosed. The NRC posts all comment submissions at http://www.regulations.gov as well as enters the comment submissions into ADAMS. The NRC does not edit comment submissions to remove identifying or contact information.

If you are requesting or aggregating comments from other persons for submission to the NRC, then you should inform those persons not to include identifying or contact information in their comment submissions that they do not want to be publicly disclosed. Your request should state that the NRC will not edit comment submissions to remove such information before making the comment submissions available to the public or entering the comment submissions into ADAMS.

Specific Comments

These comments will be submitted to the review process. Submitted to Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2012-0192.

COMMENT 1 -- Your Comment Tracking Number: 8111ad8e

The primary change to the license is to move surveillance frequencies from the license document to a document under the control of the operator. This change would only be entertained if it were the desire of the operator to decrease the frequency of surveillance to save money. The license has no upper limit on the frequency of surveillance... the operator can check these parameter MORE frequently without any concern, and no violation of the license as it stands. COPS objects to these changes for the following reasons.

We see two categories of surveillances:

1. Measurements of critical operational parameters to allow the reactor to continue to operate.
2. Tests of backup and safety equipment that is not necessary for the normal operation of the plant but are standing ready in case an emergency might unfold.

It is our observation that surveillance frequencies of critical operational parameters (1 above) are far too low (infrequent) to allow operators to -- through those surveillances -- catch an ongoing failure of the plant. For example, checking leakage from the steam generators only once every 72 hours is ridiculously low. A leak can progress quickly within only a matter of hours during a SGTR, and if the operator waits for 72 hours to detect that failure, the plant will certainly be experiencing a full Loss of cooling Accident (LOCA).

On the other hand, systems in category 2 above can be reasonably placed in a secondary document since it is only necessary to check that these systems have not deteriorated due to time, corrosion, lack of maintenance, etc. and are not involved in the critical normal operation of the plant.

Transferring these surveillance frequencies completely to the document under the control of the operator gives them too much free reign, akin to writing a purchase order to a vendor with no limit on price. A responsible way to handle this would be to include limits to the surveillance frequencies to insure that they are inspected at least more frequently than X (analogous to a Not To Exceed limit).

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> Classify all surveillances according to whether they are in class 1 or 2.

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> Increase substantially the surveillance frequencies in Class 1 to reflect the need to detect rapid deterioration in a SGTR, for example. These should not be moved to the Surveillance Frequency program.

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> Include minimum frequencies (or maximal time intervals between inspections) in the license document to insure that the licensee adheres to a reasonable limits for inspections of parameters in CLASS 2.

COMMENT 2 -- Comment Tracking Number: 8111ad90

• Attachment 1 Vol 7 (Chapter 3.4 Reactor Coolant System (RCS)) - ML 11251 A 100
  • Page 99, the proposed change is to reduce SG level from 25% to 20%. this significantly reduces the level for reactor trip. Proposal under consideration is to change 25% to 20% in two places here. This is the reverse of most changes that go from 25% to 50%, and may be a mistake. Perhaps 20% should be 50%.

The text is:

Each OPERABLE loop consists of two RCPs providing forced flow for heat transport to an SG that is OPERABLE. SG, and hence RCS loop, OPERABILITY with regard to SG water level is ensured by the Reactor Protection System (RPS) in MODES 1 and 2. A reactor trip places the plant in MODE 3 if any SG level is ≤ [25]% as sensed by the RPS. The minimum water level to declare the SG OPERABLE is [25]%.

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> This proposed change is a reduction of the level of water in the steam generator to allow the reactor to run. COPS objects to this loosening of licensee requirement and puts the plant in severe danger.

COMMENT 3 -- Comment Tracking Number: 8111ad92

COPS is concerned that that the surveillance frequency is too infrequent for checking the status of critical operational measurements to account for the rapid response needed in a real failure event.

• Attachment 1 Vol 7 (Chapter 3.4 Reactor Coolant System (RCS)) - ML 11251 A 100
  • Page 351 - CTS SR 3.4.13.2 requires verifying that primary to secondary LEAKAGE is ≤ 150 gallons per day through any one SG every 72 hours. ITS SRs 3.4.13.1 and 3.4.13.2 require similar surveillances and specify the periodic Frequencies as "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequency for the SR and the Bases for the Frequency to the Surveillance Frequency Control Program.

---

> We assert that this spec (operator much check leakage rate every 72 hours) is far too lax because the leakage can progress from a small leak to a major SGTR in only an hour or two. The SONGS steam generator leak started at a 75 gpd rate and within one hour had increased to 104 gpd. Waiting 72 hours would allow this to progress to a full SGTR and perhaps LOCA.

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> Waiting for this leak to progress to 150 gal/day rate is far too lax to detect a dangerous operating condition in the plant.

COMMENT 4 -- Comment Tracking Number: 8111ad94

The operational license has a severe internal inconsistency. On one hand, it says there can be no pressure boundary leakage at all, due to material degradation.

• From Technical Specification:

No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Definition from 10CFR 50.2 (definitions)

Reactor coolant pressure boundary means all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves, which are:

(1) Part of the reactor coolant system, or

(2) Connected to the reactor coolant system, up to and including any and all of the following:

(i) The outermost containment isolation valve in system piping which penetrates primary reactor containment,

(ii) The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment,

(iii) The reactor coolant system safety and relief valves.

For nuclear power reactors of the direct cycle boiling water type, the reactor coolant system extends to and includes the outermost containment isolation valve in the main steam and feedwater piping.

But then, it allows significant leakage to occur, up to 150 gallons per day through any one SG, and they have to check for this.

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> The definition of pressure boundary or the technical specification regarding leakage must be revised to achieve internal consistency. Now, the document is inconsistent because it first says no leakage is allowed, and then it allow leakage of up to 150 gal/day which is then released into the environment.

COMMENT 5 -- Comment Tracking Number: 8111add6

• Attachment 1 Vol 10 (Chapter 3.7 Plant Systems) - ML 11251 A 103
  • Page 99 - ADV - Atmospheric Dump Valve - The ISTS LCO 3.7.4 is being changed from "Two ADV lines shall be OPERABLE" to "One ADV line per required steam generator shall be OPERABLE." The ISTS is written such that there are two ADV lines per SG. SONGS has just one ADV line per SG and in MODE 4 SONGS could have one SG being utilized for heat removal. If the LCO required two ADV lines to be OPERABLE, SONGS would be in an ACTION unnecessarily. Therefore, the LCO was changed to require one ADV line per required steam generator. Also, due to SONGS just having one ADV line per steam generator, the Completion Time for ACTION A was changed from 7 days to 72 hours. These changes are also consistent with the SONGS Units 2 and 3 CTS.

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> We object to this design deficiency in the SONGS plant. This points out a design deficiency of SONGS compared with other plants.

• • Page 101: This part was deleted: "Two ADV lines per steam generator are required to meet single failure assumptions following an event rendering one steam generator unavailable for Reactor Coolant System (RCS) heat removal."
  • Page 102: "The design must accommodate the single failure of one ADV to open on demand; (following deleted:) thus, each steam generator must have at least two ADVs. (end delete)

---

> Since the design must accommodate the single failure of one ADV, how is this accomplished if there is only one ADV per SG??

---

> We object to this change to the license which incorrectly allows a single ADV.

COMMENT 6 -- Comment Tracking Number: 8111add8

• • Page 361 of 554, "The USFAR (ref. 3) analysis for SGTR assumes the contaminated secondary fluid reaching each SG is released via the safety valves, and by the atmospheric dump valve used to perform the plant cooldown to shutdown cooling entry. The 0.5 gpm primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential. The SLB (steam line break) is more limiting for site radiation releases. The safety analysis for the SLB accident assumes 0.5 gpm primary to secondary LEAKAGE to each steam generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 50.67"

10 CFR 50.67 (excerpt)

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

10 CFR 50.02 (excerpt)

Exclusion area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area shall normally be prohibited. In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.

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> The exclusion zone at San Onofre Nuclear Generating Station includes a freeway and an accessible beach. There are no signs warning people that ingress to the area may subject them to higher than specified radiation in the event of a rapid emergency situation.

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> Contrary to the definition of an exclusion zone, there is no means to stop traffic on the freeway in the event of a SGTR or LOCA, events that can progress within minutes and may require the complete shutdown of the freeway. Licensee should be required to install gates and turn-arounds to allow that traffic be completely stopped on the freeway and rerouted to other roads.

See this document which shows the exclusion zone on a satellite view of the facility.

• SanOnofreExclusionZone.pdf: San Onofre Exclusion Zone Analysis

Detailed Review

The changes are analyzed one by one and are classified into one of five categories.

• Administrative changes (A) - Editorial (reformatting, renumbering, and rewording) Changes to the CTS that do not result in new requirements or change operational restrictions or flexibility. These changes are supported in aggregate by a single generic no significant hazards consideration (NSHC)
  • These are probably not a concern to review by COPS
• More restrictive changes (M)—Changes to the CTS that result in added restrictions or reduced flexibility. These changes include additional requirements that decrease allowed outage times, increase the Frequency of Surveillances, impose additional Surveillances, increase the scope of Specifications to include additional plant equipment, increase the Applicability of Specifications, or provide additional actions. These changes are generally made to conform to NUREG-1432 and have been evaluated to not be detrimental to plant safety.
  • These may be of interest to review to see what they are.
• Relocated specifications (R)—Changes to the CTS that relocate CTS LCOs to licensee-controlled documents. (specifications that do not meet the selection criteria of Title 10 of the Code of Federal Regulations (10 CFR) 50.36(c)(2)(ii)). The affected structures, systems, components or variables are not assumed to be initiators of analyzed events and are not assumed to mitigate accident or transient events.
  • May be a concern to COPS
• Removed detail changes (LA)—Changes to the CTS that eliminate detail and relocate the detail to a licensee-controlled document. Typically, this involves details of system design and function, or procedural detail on methods of conducting a Surveillance Requirement (SR). Some of the proposed changes involve moving details out of the CTS and into the TS Bases, the UFSAR, the Containment Leakage Rate Testing (CLRT) Program, the LCS, or other documents under regulatory control, such as the Offsite Dose Calculation Manual (ODCM), the Quality Assurance Program (QAP), the Inservice Testing (IST) Program, the Inservice Inspection (ISI) Program, and the Surveillance Frequency Control Program (SFCP). The removal of this information is considered to be less restrictive because it is no longer controlled by the TS change process.
  • These should be reviewed and may be a concern since they are moving control of an issue to the licensee. * Less restrictive changes (L)—Changes to the CTS that result in reduced restrictions or added flexibility. These changes are supported either in aggregate by a generic NSHC that addresses a particular category of less restrictive change, or by a specific NSHC if the change does not fall into one of the eight categories of less restrictive changes. The eight categories of less restrictive changes are designated as:
  • Category 1 - Relaxation of LCO Requirements
  • Category 2—Relaxation of Applicability
  • Category 3—Relaxation of Completion Time
  • Category 4—Relaxation of Required Action
  • Category 5—Deletion of Surveillance Requirement
  • Category 6—Relaxation of Surveillance Requirement Acceptance Criteria
  • Category 7—Relaxation of Surveillance Frequency
  • Category 8—Deletion of Reporting Requirements

Source Documents:

Best way to get these is through ADAMS, with Document Content = "San Onofre" and Document Date = 07/29/2011. Some of the links provided here do not work for some reason. Or, search for the document number by each one.

• https://www.federalregister.gov/articles/2012/08/16/2012-20114/southern-california-edison-san-onofre-nuclear-generating-station-units-2-and-3-application-and
• CFR 50.36 on Technical Specifications: http://www.gpo.gov/fdsys/pkg/CFR-2011-title10-vol1/xml/CFR-2011-title10-vol1-sec50-36.xml
• Main application for LAR: ML11251A113
• Attachment 1 Vol 1 (Chapter 1.0 Use and Application) - ML11251A094
• Attachment 1 Vol 2 (Chapter 2.0 Safety Limits (SLs)) - ML11251A095
• Attachment 1 Vol 3 (Chapter 3.0 Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability) - ML11251A096
• Attachment 1 Vol 4 (Chapter 3.1 Reactivity Control Systems) - ML11251A108
• Attachment 1 Vol 5 (Chapter 3.2 Power Distribution Limits) - ML11251A098
• Attachment 1 Vol 6 (Chapter 3.3 Instrumentation) - ML11251A099
• Attachment 1 Vol 7 (Chapter 3.4 Reactor Coolant System (RCS)) - ML11251A100
  • Page 99, reduce SG level from 25% to 20%. this seems like a significant change. Proposal is to change 25% to 20% in two places here. This is the reverse of most changes that go from 25% to 50%, and may be a mistake. Perhaps 20% should be 50%.

Each OPERABLE loop consists of two RCPs providing forced flow for heat transport to an SG that is OPERABLE. SG, and hence RCS loop, OPERABILITY with regard to SG water level is ensured by the Reactor Protection System (RPS) in MODES 1 and 2. A reactor trip places the plant in MODE 3 if any SG level is ≤ [25]% as sensed by the RPS. The minimum water level to declare the SG OPERABLE is [25]%.

• • page 124,128,155,162,190,196,198,200 - More stringent requirement for >= 50% instead of 25%, but sometimes add "(wide range)" and I'm not sure what that means.
  • Page 138,143,179 - delete "wide range"
  • Page 190,192 says > 50% instead of >= 50%

• • Page 351 - CTS SR 3.4.13.2 requires verifying that primary to secondary LEAKAGE is ≤ 150 gallons per day through any one SG every 72 hours. ITS SRs 3.4.13.1 and 3.4.13.2 require similar surveillances and specify the periodic Frequencies as "In accordance with the Surveillance Frequency Control Program." This changes the CTS by moving the specified Frequency for the SR and the Bases for the Frequency to the Surveillance Frequency Control Program.
    • The relocation of the specified Surveillance Frequencies to licensee control is the primary change in this License Amendment Request. The surveillance rates may be the same now, but if they are under licensee control, we can expect them to gut the surveillance rates.

The relocation of the specified Surveillance Frequencies to licensee control is consistent with Regulatory Guides 1.174 and 1.177. Regulatory Guide 1.177 provides guidance for changing Surveillance Frequencies and Completion Times. However, for allowable risk changes associated with Surveillance Frequency extensions, it refers to Regulatory Guide 1.174, which provides quantitative risk acceptance guidelines for changes to core damage frequency (CDF) and large early release frequency (LERF). Regulatory Guide 1.174 provides additional guidelines that have been adapted in the risk-informed methodology for controlling changes to Surveillance Frequencies.

• • • COPS: This spec (operator much check leakage rate every 72 hours) is far too lax because the leakage can progress from a small leak to a major SGTR in only an hour or two. The SONGS steam generator leak started at a 75 gpd rate and within an hour had increased to 104 gpd. Waiting 72 hours would allow this to progress to a full SGTR and perhaps LOCA.
    • See "Risk-Informed Method for Control of Surveillance Frequencies" http://pbadupws.nrc.gov/docs/ML0713/ML071360456.pdf

• • move surveillance times to separate document -- makes it harder to find the times?
  • Spec changed from "Average Temp of RCS is >= 520 F" to "Cold leg is >= 522 F"
  • SR 3.4.5.2 Verify secondary side water level in each steam generator was 25% change to 50%. Frequently, it uses the wording like this: "This SR requires verification (per surveillance plan) that the secondary side water level in each SG is >= 50% (wide range). An adequate SG water level is required in order to have a heat sink for removal of the core decay heat from the reactor coolant"

(Type 3 – Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS SR 3.4.6.2 requires verifying that secondary side water level in each steam generator ≥ 50% (wide range). ITS SR 3.4.6.2 requires verifying that secondary side water level in each steam generator ≥ 50%. This changes the CTS by deleting the parenthetical statement, "(wide range)," from the Surveillance Requirement to verify secondary side water level is ≥ 50%.

The removal of the parenthetical statement, "(wide range)," from CTS SR 3.4.6.2 is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS Bases for SR 3.4.6.2 contains the information that the steam generator level indication is ≥ 50% wide range. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details that require level indication to be wide range is being moved from the Technical Specifications to the ITS Bases.

• • Page 361 of 554, "The USFAR (ref. 3) analysis for SGTR assumes the contaminated secondary fluid reaching each SG is released via the safety valves, and by the atmospheric dump valve used to perform the plant cooldown to shutdown cooling entry. The 0.5 gpm primary to secondary LEAKAGE safety analysis assumption is relatively inconsequential. The SLB (steam line break) is more limiting for site radiation releases. The safety analysis for the SLB accident assumes 0.5 gpm primary to secondary LEAKAGE to each steam generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 50.67"

10 CFR 50.67 (excerpt)

The NRC may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

• From Technical Specification:

No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Definition from 10CFR 50.2 (definitions)

Reactor coolant pressure boundary means all those pressure-containing components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves, which are:

(1) Part of the reactor coolant system, or

(2) Connected to the reactor coolant system, up to and including any and all of the following:

(i) The outermost containment isolation valve in system piping which penetrates primary reactor containment,

(ii) The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment,

(iii) The reactor coolant system safety and relief valves.

For nuclear power reactors of the direct cycle boiling water type, the reactor coolant system extends to and includes the outermost containment isolation valve in the main steam and feedwater piping.

• • Page 364 - (FYI) "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

• • 3.4.17 Steam Generator (SG) tube integrity -- See pages 499+
    • Change "plug or repaired" to just "plug"

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.9, "Steam Generator Program," and describe acceptable SG tube performance

• • Page 510 - Describes the SGTR accident. Specs are to avoid a SGTR.

• • Page 510 - This paragraph doesn't make much sense. Why would any tube that satisfies the repair criteria not be plugged. Plus "Repair Criteria" should be "Plug Criteria"

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

• Attachment 1 Vol 8 (Chapter 3.5 Emergency Core Cooling Systems (ECCS)) - ML11251A101
• Attachment 1 Vol 9 (Chapter 3.6 Containment Systems) - ML11251A102
• Attachment 1 Vol 10 (Chapter 3.7 Plant Systems) - ML11251A103
  • Page 5 - Main Steam Safety Valves (MSSVs) - change action condition from two to seven inoperable to one or more inoperable.. (Better).
  • Page 99 - ADV - Atmospheric Dump Valve - The ISTS LCO 3.7.4 is being changed from "Two ADV lines shall be OPERABLE" to "One ADV line per required steam generator shall be OPERABLE." The ISTS is written such that there are two ADV lines per SG. SONGS has just one ADV line per SG and in MODE 4 SONGS could have one SG being utilized for heat removal. If the LCO required two ADV lines to be OPERABLE, SONGS would be in an ACTION unnecessarily. Therefore, the LCO was changed to require one ADV line per required steam generator. Also, due to SONGS just having one ADV line per steam generator, the Completion Time for ACTION A was changed from 7 days to 72 hours. These changes are also consistent with the SONGS Units 2 and 3 CTS.
    • COPS: This points out a design deficiency of SONGS compared with other plants, apparently.
  • Page 101: This part was deleted: "Two ADV lines per steam generator are required to meet single failure assumptions following an event rendering one steam generator unavailable for Reactor Coolant System (RCS) heat removal."
  • Page 102: "The design must accommodate the single failure of one ADV to open on demand; (following deleted:) thus, each steam generator must have at least two ADVs. (end delete)
    • COPS: Since the design must accommodate the single failure of one ADV, how is this accomplished if there is only one ADV per SG??

• Attachment 1 Vol 11 (Chapter 3.8 Electrical Power Systems) - ML11251A105
• Attachment 1 Vol 12 (Chapter 3.9 Refueling Operations) - ML11251A106
• Attachment 1 Vol 13 (Chapter 4.0 Design Features) - ML11251A107
• Attachment 1 Vol 14 (Chapter 5.0 Administrative Controls) - ML11251A108
• Attachment 1 Vol 15 (Generic NSHCs) - ML11251A109

Notes:

As usual, obfuscation starts with a whole new language, and a plethora of acronyms.

• LAR = license amendment request
• RAI = requests for additional information - NRC staff issued or will issue its requests for additional information (RAIs) and the licensee addressed or will address the RAIs through the ITS Conversion Web page.
• DOCs = discussion of changes
• NSHC = no significant hazards consideration -- this is a document or section of a document that argues that a change can be made without any hazard or safety concerns.
• LCO = Limiting Conditions for Operation -- defined in the CFR regarding Technical Specifications

• AFW = AUXILIARY FEEDWATER
• CCW = Component Cooling Water system
• MSSVs = Main Steam Safety Valves

• RCS = Reactor Cooling System -- An OPERABLE RCS loop consists of at least one OPERABLE RCP and an SG that is OPERABLE and has the minimum water level specified in SR 3.4.6.2. [(50%)??]

• SDC train = Shutdown Cooling Train -- an OPERABLE SDC train is composed of the OPERABLE SDC pump(s) capable of providing forced flow to the SDC heat exchanger(s). RCPs and SDC pumps are OPERABLE if they are capable of being powered and are able to provide flow if required.

• SLB = steam line break -- The SLB radiological analysis assumes that offsite power is lost at the same time as the pipe break occurs outside containment. The affected SG blows down completely and steam is vented directly to the atmosphere. The unaffected SG removes core decay heat by venting steam to the atmosphere until the cooldown ends and the SDC system is placed in service.
• SGTR = steam generator tube rupture -- (pg 484) The SGTR analysis assumes the contaminated secondary fluid reaching each SG is released via the safety valves, and by atmospheric dump valve used to perform the plant cooldown to shutdown cooling entry. The unaffected SD removes core decay heat by venting steam until the colldown ends adn the Shutdown Cooling (SDC) System is placed in service.
• SSC = structures, systems and components
• SFCP = Surveillance Frequency Control Program
• STIs = surveillance test intervals
• ALARA = as low as reasonably achievable
• PRA = probability Risk Assessment
• CDF = core damage frequency - Used in PRA to determine if a system should be tested, i.e. given surveillance.
• LERF = large early release frequency - Also used in PRA.

CFR 50.36 Technical specifications. (c)

(2) Limiting conditions for operation.

(i) Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. When a limiting condition for operation of any process step in the system of a fuel reprocessing plant is not met, the licensee shall shut down that part of the operation or follow any remedial action permitted by the technical specifications until the condition can be met. In the case of a nuclear reactor not licensed under § 50.21(b) or § 50.22 of this part or fuel reprocessing plant, the licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the nuclear reactor or the fuel reprocessing plant. In the case of nuclear power reactors licensed under § 50.21(b) or § 50.22, the licensee shall notify the Commission if required by § 50.72 and shall submit a Licensee Event Report to the Commission as required by § 50.73. In this case, licensees shall retain records associated with preparation of a Licensee Event Report for a period of three years following issuance of the report. For events which do not require a Licensee Event Report, the licensee shall retain each record as required by the technical specifications.

(ii) A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

(iii) A licensee is not required to propose to modify technical specifications that are included in any license issued before August 18, 1995, to satisfy the criteria in paragraph (c)(2)(ii) of this section.

Process to request or petition for a hearing

The following excerpt explains this process.

All documents filed in NRC adjudicatory proceedings, including a request for hearing, a petition for leave to intervene, any motion or other document filed in the proceeding prior to the submission of a request for hearing or petition to intervene, and documents filed by interested governmental entities participating under 10 CFR 2.315(c), must be filed in accordance with the NRC E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to submit and serve all adjudicatory documents over the internet, or in some cases to mail copies on electronic storage media. Participants may not submit paper copies of their filings unless they seek an exemption in accordance with the procedures described below.

To comply with the procedural requirements of E-Filing, at least ten 10 days prior to the filing deadline, the participant should contact the Office of the Secretary by email at hearing.docket@nrc.gov, or by telephone at 301-415-1677, to request (1) a digital identification (ID) certificate, which allows the participant (or its counsel or representative) to digitally sign documents and access the E-Submittal server for any proceeding in which it is participating; and (2) advise the Secretary that the participant will be submitting a request or petition for hearing (even in instances in which the participant, or its counsel or representative, already holds an NRC-issued digital ID certificate). Based upon this information, the Secretary will establish an electronic docket for the hearing in this proceeding if the Secretary has not already established an electronic docket.

Information about applying for a digital ID certificate is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. System requirements for accessing the E-Submittal server are detailed in the NRC's “Guidance for Electronic Submission,” which is available on the NRC's public Web site at http://www.nrc.gov/site-help/e-submittals.html. Participants may attempt to use other software not listed on the Web site, but should note that the NRC's E-Filing system does not support unlisted software, and the NRC Meta System Help Desk will not be able to offer assistance in using unlisted software.

From the "apply-certificates" page:

There are three steps to obtain a Veri Sign Digital ID Certificate from NRC:

• Step 1 - Request an NRC Approval Code
  • To request an NRC Approval Code you must first determine which NRC program meets your needs. Each program area will want basic information from you including your name, email address, phone number, organization, role in the organization and reason for submitting documents to the NRC.
  • Adjudicatory Proceedings Program participants should call (301) 415-1677 or send an email request to the NRC’s Office of the Secretary staff at: Hearing.Docket@nrc.gov
• Step 2 - Enroll for your Digital ID Certificate --
  • http://www.nrc.gov/site-help/e-submittals/certificate/enroll-certificate.html
  • VIew and print this document: http://www.nrc.gov/site-help/e-submittals/certificate/nrc-digital-ids-guide.pdf
  • Pick up your digital ID certificate from the same PC where you enrolled. Only enroll from the PC where you receive email.
  • You must have an NRC issued digital ID Approval Code prior to electronic enrollment for a certificate.
• Step 3 - Install and Use a Digital ID Certificate

Interaction with NRC

From: Ray Lutz [mailto:raylutz@citizensoversight.org]
Sent: Thursday, September 20, 2012 7:33 PM
To: Docket, Hearing
Subject: Request for NRC Approval Code


To: NRC's Office of the Secretary
Re: Request for NRC Approval Code for the Adjudicatory Proceedings Program

Dear Representative:

I hereby request an NRC Approval Code for the Adjudicatory Proceedings
Program.

My Name:         Ray Lutz
Email:                 raylutz@citizensoversight.org
Phone:               619-447-3246 or 619-820-5321 (mobile)
Organization:    Citizens Oversight Projects (COPS) -- (Citizens
Oversight, Inc)
Role:                   Engineer -- Technical Analyst

Reason for submitting documents:

It is our intention to request a hearing and/or petition for leave to
intervene, regarding Docket ID NRC-2012-0192 "Southern California
Edison, San Onofre Nuclear Generating Station, Units 2 and 3;
Application and Amendment to Facility Operating License Involving
Proposed No Significant Hazards Consideration Determination"

--Ray Lutz

On 9/21/2012 6:05 AM, Docket, Hearing wrote:
Mr. Lutz:

Good morning.  Please let me know if you are referring to a recent Federal Register Notice and if possible the date that it was published?

Thank you,


Rebecca Giitter
Rulemakings and Adjudications Staff
Office of the Secretary
U.S. Nuclear Regulatory Commission
(301) 415-1679

From: Ray Lutz [mailto:raylutz@citizensoversight.org] 
Sent: Sunday, September 23, 2012 12:17 AM
To: Docket, Hearing
Subject: Re: Request for additional information

Yes, it is, here is the link to the notice:
https://www.federalregister.gov/articles/2012/08/16/2012-20114/southern-california-edison-san-onofre-nuclear-generating-station-units-2-and-3-application-and

ENTITLED:
Southern California Edison, San Onofre Nuclear Generating Station, Units 
2 and 3; Application and Amendment to Facility Operating License 
Involving Proposed No Significant Hazards Consideration Determination

A Notice by the Nuclear Regulatory Commission on 08/16/2012

-- Ray Lutz

Mr. Lutz:

Good morning.  Thank you for your response.  This is a new proceeding and I will need to establish a new docket for it.  I will send you an email when the docket has been established.  

Thank you,

Rebecca Giitter
Rulemakings and Adjudications Staff
Office of the Secretary
U.S. Nuclear Regulatory Commission
(301) 415-1679

Mr. Lutz:

The Office of the Secretary has received your request and is providing you with an approval code to apply online for a new digital ID (Veri Sign Certificate). You will need to apply for your digital certificate through the NRC Electronic Information Exchange (EIE) website. See the instructions at http://www.nrc.gov/site-help/e-submittals/apply-certificates.html. Please use the following approval code =ART09RLCI= when completing the online form. After you have completed the enrollment page, you will receive 2 separate emails from the digital ID administrator stating that your request is being processed and again to notify that your digital certificate is ready to be picked up and downloaded.

(NEW LINK) Digital ID Center https://onsite.verisign.com/services/USNuclearregulatoryCommission/digitalidCenter.html

Once you have downloaded your digital certificate it will be necessary to request access to a proceeding and you can do so by using this link https://eieprod.nrc.gov/EIE25

A window will appear and you will need to select your digital certificate and select ok. At the EIE Terms of Service page click on "Accept" to proceed. At El E Portal page under “EIE User Maintenance”, click on "Request Access to a Proceeding." Complete the required information and click on "Submit Request" at the bottom of the page. You will receive an automated email message stating that you have been granted access to the requested proceeding.

Related Links This page contains written guidance and video's designed to assist you in the digital certificate application and download process. http://www.nrc.gov/site-help/e-submittals/apply-certificates.html

Additional requirements

You must be familiar with the document format requirements for submitting an EIE filing. See http://www.nrc.gov/site-help/e-submittals.html to obtain submission guidance and view the instructional video:

In order for you to access and view documents submitted via EIE you must download and install the Workplace Forms Viewer. For guidance see: http://www.nrc.gov/site-help/e-submittals/install-viewer.html

If you require assistance with installation of your Certificate or Forms Viewer, with requesting access, or with submission of documents via our electronic filing system, please contact our Help Desk toll-free at 1-866-672-7640 or MSHD.Resource@nrc.gov. They are available Mondays through Fridays, except federal holidays, during the hours of 8:00 a.m. to 8:00 p.m. Eastern time.

Note that a request for a digital certificate must come directly from the individual.

Thank you.

 
Rebecca Giitter
Rulemakings and Adjudications Staff
Office of the Secretary
U.S. Nuclear Regulatory Commission
(301) 415-1679

Completed the enrollment form and received this response:

Your Digital Certificate request has been submitted for approval

Once your digital certificate request has been approved, you will receive an email with instructions for installing your digital certificate. If you do not receive it shortly, contact your administrator.

Dear Ray Lutz,

Thank you for requesting a Digital ID. The U.S. Nuclear Regulatory Commission (NRC) is processing your request, and will notify you when your Digital ID is ready. The NRC approval process should happen fairly quickly but sometimes takes one or two days.

If you have questions about your application or haven't received a reply from the NRC within 2 business days, please contact the Digital Certificate Administrator at NRC by replying to this email message or by calling the NRC's Digital ID Help Line at (301) 415-0439.

Received 2012-09-27 at 6:00 AM

Mr. Lutz:

Good morning. The docket for the San Onofre 50-361 and 50-362-LA has been established. Once you have downloaded your digital certificate it will be necessary to request access to a proceeding and you can do so by using this link https://eieprod.nrc.gov/EIE25

A window will appear and you will need to select your digital certificate and select ok. At the EIE Terms of Service page click on "Accept" to proceed. At El E Portal page under "EIE User Maintenance", click on "Request Access to a Proceeding." Complete the required information and click on "Submit Request" at the bottom of the page. You will receive an automated email message stating that you have been granted access to the requested proceeding.

Additional requirements

You must be familiar with the document format requirements for submitting an EIE filing. See http://www.nrc.gov/site-help/e-submittals.html to obtain submission guidance and view the instructional video:

In order for you to access and view documents submitted via EIE you must download and install the Workplace Forms Viewer. For guidance see: http://www.nrc.gov/site-help/e-submittals/install-viewer.html

If you require assistance with installation of your Certificate or Forms Viewer, with requesting access, or with submission of documents via our electronic filing system, please contact our Help Desk toll-free at 1-866-672-7640 or MSHD.Resource@nrc.gov. They are available Mondays through Fridays, except federal holidays, during the hours of 8:00 a.m. to 8:00 p.m. Eastern time.

Rebecca Giitter
Rulemakings and Adjudications Staff
Office of the Secretary
U.S. Nuclear Regulatory Commission
(301) 415-1679

Received 2012-09-27 at 1:59pm

Congratulations!

Your Digital ID has been successfully generated and installed.

Your Digital ID Information.

Organization = U.S. Nuclear regulatory Commission
Organizational Unit = www.verisign.com/repository/CPS Incorp. by Ref.,LIAB.LTD(c)99
Title = Engineer & National Coordinator
Common Name = Ray Lutz
Email Address = raylutz@citizensoversight.org
Serial Number = 05a99bb0c115168957b90ffc1230c497

---

Installation of certificate failed, not sure why. Had to call technical support and had to go through the whole process again, from the beginning.

PIN = 9714547888

Organization = U.S. Nuclear regulatory Commission
Organizational Unit = www.verisign.com/repository/CPS Incorp. by Ref.,LIAB.LTD(c)99
Title = Engineer / National Coordinator
Common Name = Ray Lutz
Email Address = raylutz@citizensoversight.org
Serial Number = 12d0f94dcb712cbe92a765bed5977809

Then, went to this Electronic Information Exchange Portal: https://eieprod.nrc.gov/EIE25/portal.do and requested access to the proceeding that was already set up. Made the request and the contact at the help desk said it will take a couple of days for them to approval it.

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