ML13268A143

Request for Exemption from Requirements of 10 CFR 50.54(m) and 10 CFR 55

ML13268A143
Person / Time
Site:	San Onofre
Issue date:	09/11/2013
From:	St.Onge R Southern California Edison Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML13268A143 (104)

Text

SOUTHERN CALIFORNIA Richard St. Onge

,] EDISON Director, Nuclear Regulatory Affairs and EO Emergency Planning An EDISON INTERNATIONALU Company 10 CFR 50.54(m) 10 CFR 55 10 CFR 50.12 September 11, 2013 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Request for Exemption from Requirements of 10 CFR 50.54(m) and 10 CFR 55 Docket Nos. 50-361 and 50-362 San Onofre Nuclear Generating Station, Units 2 and 3

Reference:

1. Letter from P. T. Dietrich (SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated June 28, 2013;

Subject:

Permanent Removal of Fuel from the Reactor Vessel, San Onofre Nuclear Generating Station Unit 3

2. Letter from P. T. Dietrich (SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated July 22, 2013;

Subject:

Permanent Removal of Fuel from the Reactor Vessel, San Onofre Nuclear Generating Station Unit 2

3. Letter from R. St. Onge (SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated August 20, 2013;

Subject:

Request for Approval of the Safe Storage Shift Manager/Certified Fuel Handler Training Program, San Onofre Nuclear Generating Station Units 2 and 3

Dear Sir or Madam:

Pursuant to 10 CFR 50.12, Southern California Edison (SCE) requests exemptions from 10 CFR 50.54(m) and 10 CFR 55 for San Onofre Nuclear Generating Station (SONGS),

Units 2 and 3. The proposed exemptions would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS Units 2 and 3, with certified fuel handlers and non-licensed operators, to conform to the permanently defueled condition of the station.

By letters dated June 28, 2013 and July 22, 2013, SCE submitted certifications of permanent removal of fuel from the reactor vessels for SONGS Units 3 and 2 (References 1 and 2). Consequently, the 10 CFR 50 licenses for SCE no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor P.O. Box 128 San Clemente, CA 92674 0

[j -L

Document Control Desk September 11, 2013 vessel, as specified in 10 CFR 50.82(a)(2). On August 20, 2013, SCE submitted a request for Commission approval of a certified fuel handler training program (Reference 3).

10 CFR 50.54(m) specifies staffing levels for operators licensed pursuant to 10 CFR 55, "Operators' Licenses." Based on the scope of 10 CFR 55, which requires licensed individuals only for the manipulation of the controls of a utilization facility, SCE no longer requires operators licensed under this part because it no longer has "controls", as defined by 10 CFR 55.4. SCE's understanding is that 10 CFR 50.54(m) is inapplicable since this regulation serves to specify staffing levels for operators licensed and, therefore, no longer has "operators" as defined by 10 CFR 55.4. Recently, NRC staff verbally informed SCE of their interpretation that 10 CFR 50.54(m) continues to apply to facilities that have submitted a certification of permanent removal of fuel from the reactor vessel but have not received a license amendment addressing a change in shift staffing.

To address this NRC concern, exemptions from 10 CFR 50.54(m) and 10 CFR 55 are requested, The requested exemptions from 10 CFR 50.54(m) and 10 CFR 55 are permissible under 10 CFR 50.12 because it will not present an undue risk to the public health and safety, and application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rule. Therefore, application of the subject portions of 10 CFR 50.54(m) is not necessary to ensure adequate operator staffing levels at SONGS. Application of the rule would impose an unnecessary burden on SCE to maintain a licensed operator training program.

The exemption request is contained in the enclosure to this letter. SCE requests approval of this exemption request on an expedited basis, and no later than November 30, 2013. Correspondingly, SCE also requests expedited approval of the certified fuel handler training program (Reference 3).

This letter does not contain any commitments.

If you have any questions regarding this matter, please contact Mr. M. E. Morgan at (949) 368-6745.

Sincerely,

Document Control Desk September 11, 2013

Enclosures:

1. Request for Exemptions from 10 CFR 50.54(m) and 10 CFR 55

2. Revised Updated Final Safety Analysis "Chapter 15 Accident Analysis" Sections 15.0 and 15.7

3. Procedure S0123-0-Al "Conduct of Operations" Revision 41 cc:

S. A. Reynolds, Regional Administrator (Acting), NRC Region IV B. J. Benney, NRC Project Manager, San Onofre Units 2 & 3 G. G. Warnick, NRC Senior Resident Inspector, San Onofre Units 2 & 3

ENCLOSURE1 Request for Exemptions from 10 CFR 50.54(m) and 10 CFR 55 Pursuant to 10 CFR 50.12, Southern California Edison (SCE) requests exemptions from 10 CFR 50.54(m) and 10 CFR 55 for San Onofre Nuclear Generating Station (SONGS),

Units 2 and 3. The proposed exemptions would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS Units 2 and 3, with certified fuel handlers and non-licensed operators, to conform to the permanently defueled condition of the station.

I.

DESCRIPTION Pursuant to 10 CFR 50.12, Southern California Edison (SCE) requests exemptions from 10 CFR 50.54(m) and 10 CFR 55 for San Onofre Nuclear Generating Station (SONGS).

10 CFR 50.54(m) specifies minimum requirements per shift for on-site staffing of nuclear power units by operators and senior operators licensed under 10 CFR 55, "Operators' Licenses". 10 CFR 55 establishes procedures and criteria for issuance of licenses to operators and senior operators.

These exemptions are requested to allow SCE to replace reliance on senior reactor operators and reactor operators licensed pursuant to 10 CFR 55 at SONGS, with certified fuel handlers and certified operators, to conform to the permanently defueled condition of the station. Although the current 10 CFR 50 regulatory requirements (developed for operating reactors) ensure safety at the decommissioning facility, some of these requirements are unnecessary and do not substantially contribute to public safety.

By letters dated June 28, 2013 and July 22, 2013, SCE submitted certifications of permanent removal of fuel from the reactor vessels for SONGS Units 3 and 2 (References 1 and 2).

Upon docketing of the References 1 and 2 certifications, the 10 CFR 50 licenses for SONGS Units 2 and 3 no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2). On August 20, 2013, SCE submitted a request for Commission approval of a certified fuel handler training program (Reference 3).

The requested exemptions from 10 CFR 50.54(m) and 10 CFR 55 are permissible under 10 CFR 50.12 because they will not present an undue risk to the public health and safety, and application of the regulations in this particular circumstance is not necessary to achieve the underlying purpose of the rule. Therefore, application of the subject portions of 10 CFR 50.54(m) and 10 CFR 55 are not necessary to ensure 1

adequate operator staffing levels at SONGS. Application of the rule would impose an unnecessary burden on SCE to maintain a licensed operator training program.

10 CFR 50.54(m) is restated below.

(m)(1) A senior operator licensed pursuant to part 55 of this chapter shall be present at the facility or readily available on call at all times during its operation, and shall be present at the facility during initial start-up and approach to power, recovery from an unplanned or unscheduled shut-down or significant reduction in power, and refueling, or as otherwise prescribed in the facility license.

(2) Notwithstanding any other provisions of this section, by January 1, 1984, licensees of nuclear power units shall meet the following requirements:

(i) Each licensee shall meet the minimum licensed operator staffing requirements in the following table:

Minimum Requirements' Per Shift for On-Site Staffing of Nuclear Power Units by Operators and Senior Operators Licensed Under 10 CFR Part 55 One Unit Two units Three units Number of nuclear power units operating 2

None Position One control One control Two control Two control Three control room room rooms rooms rooms One Senior Operator Operator Senior Operator Operator Senior Operator Operator Senior Operator Operator 1

1 2

2 2

2 3

2 3

1 1

2 2

3 3

4 3

2 1

4 j3 1

3 2

4 3

5 4

6 Two Three 3

5 1Temporary deviations from the numbers required by this table shall be in accordance with criteria established in the unit's technical specifications.

2For the purpose of this table, a nuclear power unit is considered to be operating when it is in a mode other than cold shutdown or refueling as defined by the unit's technical specifications.

3The number of required licensed personnel when the operating nuclear power units are controlled from a common control room are two senior operators and four operators.

(ii) Each licensee shall have at its site a person holding a senior operator license for all fueled units at the site who is assigned responsibility for overall plant operation at all times there is fuel in any unit. If a single senior operator does not hold a senior operator license on all fueled units at 2

the site, then the licensee must have at the site two or more senior operators, who in combination are licensed as senior operators on all fueled units.

(iii) When a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the unit's technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. In addition to this senior operator, for each fueled nuclear power unit, a licensed operator or senior operator shall be present at the controls at all times.

(iv) Each licensee shall have present, during alteration of the core of a nuclear power unit (including fuel loading or transfer), a person holding a senior operator license or a senior operator license limited to fuel handling to directly supervise the activity and, during this time, the licensee shall not assign other duties to this person.

(3) Licensees who cannot meet the January 1, 1984 deadline must submit by October 1, 1983 a request for an extension to the Director of the Office of Nuclear Regulation and demonstrate good cause for the request.

Exemption is being requested from this section of the regulation in its entirety. For a facility containing two nuclear power units that are not operating, the above regulation would require one Senior Operator and two Operators onsite per shift, each of whom is licensed under 10 CFR 55. Per footnote 2, this section of the regulation considers a unit to be operating when it is in a mode (as defined by the unit's technical specifications) other than cold shutdown or refueling. Conversely, a nuclear power unit would be considered to be not operating when it is in a mode of cold shutdown or refueling. However, a permanently defueled facility is not in any mode. Therefore, the applicability of this requirement to a permanently defueled facility appears to lack clarity.

Because the 10 CFR 50 licenses for SONGS Units 2 & 3 no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), there will no longer be licensed operators as specified in 10 CFR 55 in the control room. SCE administratively controls shift crew staffing levels consistent with the requirements contained in Technical Specification 5.2.2 for the defueled condition. These current administrative requirements specify a minimum of three non-licensed operators.

Additionally, the numerical operator staffing level remains the same as previously required per 10 CFR 50.54(m). Only the qualification requirements will be changed so as to reflect the permanently defueled condition. The qualification requirements of the administratively required individuals are commensurate with the scope of activities needed for safe management of irradiated fuel at a permanently defueled facility.

Certified Fuel Handlers (CFHs) at SONGS are trained in accordance with a CFH Training Program that was submitted for NRC review and approval on August 20, 2013 (Reference 3). The CFH Training Program ensures that the qualifications of fuel handlers are commensurate with the tasks to be performed and the conditions requiring response. As documented in Reference 3, the CFH Training Program provides adequate confidence that appropriate training of personnel who will perform CFH duties 3

is conducted to ensure the facility is maintained in a safe and stable condition.

Reference 3 further stipulates the CFH Training Program uses the systematic approach to training (SAT).

The minimum shift staffing levels will be administratively controlled by SONGS procedure SO123-0-Al, Conduct of Operations, to ensure consistency with the proposed exemption. Although not reliant on operators licensed under 10 CFR 55, these controls will be consistent with those levels currently approved at other permanently defueled nuclear facilities in the U.S. that rely on Certified Fuel Handlers and non-licensed operators in lieu of licensed operators (e.g., Millstone 1 and Zion).

Based on the information provided above, SCE has concluded that 10 CFR 50.54(m) and 10 CFR 55 impose an unnecessary regulatory burden on a permanently defueled facility that is not needed for safe storage and management of irradiated fuel during decommissioning.

The requested exemptions would not present an undue risk to the public health and safety or prevent appropriate operator response from being performed as required.

II.

BACKGROUND Reduced Scope of Radiological Accidents at Permanently Defueled Facilities The irradiated fuel will be stored in the spent fuel pool (SFP) and in the Independent Spent Fuel Storage Installation (ISFSI) until it is shipped off site sometime in the future.

Since the reactor is permanently defueled, the SFP and its supporting systems are dedicated only to spent fuel storage. With the reactor defueled, the reactor, RCS and secondary system are no longer in operation and have no function related to the safe storage and management of irradiated fuel.

10 CFR 50.82(a)(2) specifies that the 10 CFR 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel after docketing the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel. After the termination of reactor operations at SONGS and the permanent removal of the fuel from the reactor vessel, the postulated accidents involving failure or malfunction of the reactor, RCS or secondary system are no longer applicable.

Revised dose calculations were completed to support the changes to the Updated Final Safety Analysis Report (UFSAR) Chapter 15 Accident Analysis (Reference 7). These calculations validate that the dose consequences would be within 10 CFR 50.67 and Regulatory Guide 1.183 dose limits and would be below the Environmental Protection Agency's Protective Action Guidelines. The UFSAR was revised to reflect the new analyses.

4

Licensed Operator Staffing Requirements at Nuclear Power Units 10 CFR 50.54(m) specifies minimum requirements' per shift for on-site staffing of nuclear power units by operators and senior operators licensed under 10 CFR 55. The requirements of 10 CFR 50.54(m) were published in the Federal Register on July 11, 1983 (48 FR 31611; (Reference 4)). The summary for this final rule stated that this regulation was promulgated "to require licensees of nuclear power units to provide a minimum number of licensed operators and senior operators on shift at all times to respond to normal and emergency conditions. These requirements will further assure the protection of the health and safety of the public by allowing the senior operator in charge the flexibility to move about the facility as needed while assuring that a senior operator is continuously present in the control room during unit operation."

The background for this rule stated that this rule implemented the recommendations of the NRC Action Plan Developed as a Result of the TMI-2 Accident (which provided interim shift staffing criteria in NUREG-0737 to all licensees of operating units). This section continued by stating "To ensure that all operating nuclear power units are adequately staffed with licensed personnel, the amendment will apply these NUREG-0737 criteria to all operating nuclear power units."

Although the rule does not explicitly exclude facilities that have submitted the certifications under Section 50.82(a)(1), the language implies as much, in that both the basis and the rule itself are premised on "operating nuclear power units". Whereas, the 10 CFR 50 license of a facility that has submitted the certifications under Section 50.82(a)(1) no longer authorizes operation of the reactor. This understanding appears to be reinforced by SECY-00-0145, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning" (Reference 5), which discusses portions of 10 CFR 50.54 that contain licensed operator requirements that do not apply to decommissioning plants.

The discussion in Section D of SECY-00-0145 (Staffing and Training) states the following:

A decommissioning plant is clearly not "operating" and no manipulation of controls that affect reactor reactivity or power can occur at a permanently defueled reactor. Therefore, the regulations that require specified licensed operator staffing for operating reactors are not applicable to a decommissioning plant.

SECY-00-0145 further reinforces this position in the Recommended Approach section, which recommends that the existing rule regarding licensed operator staffing at decommissioning facilities (i.e., that licensed operator staffing requirements for operating reactors are not applicable to a decommissioning plant) be clarified. The recommendation is as follows:

5

Clarify that licensed operators are not required for permanently shutdown and defueled reactors.

An NRC memorandum regarding applicability of Title 10 of the code of federal regulations to decommissioning nuclear power plants (Reference 6) also questions whether licensed operators are required for permanently shutdown and defueled reactors. The discussion regarding 10 CFR 50.54(m) states that although this section is applicable to all licenses, "the applicability of this subsection appears to apply to plant operation." It further expounds on this position by stating that "once the fuel has been permanently removed from the reactor vessel, the applicability of this regulation becomes less apparent." The implication from the explanation in this section is that this regulation applies to decommissioning units prior to the fuel having been permanently removed from the reactor vessel. This explanation is further reinforced in the discussion regarding 10 CFR 55, Operators' Licenses, which states that "the regulations of this part are deemed to be applicable to decommissioning plants as long as licensed operators are used by the decommissioning plant for staffing." Such a position would be logical, in that the regulation governing licensed operators (10 CFR 55) should remain applicable to facilities that rely on licensed operators. Conversely, once a facility no longer relies on licensed operators, 10 CFR 55 would no longer apply.

The table in 10 CFR 50.54(m)(2)(i) that specifies licensed operator staffing levels is titled "Minimum Requirements Per Shift for On-Site Staffing of Nuclear Power Units by Operators and Senior Operators Licensed Under 10 CFR Part 55." This title indicates that the requirements contained therein specify staffing requirements for facilities that rely on licensed operators.

Per footnote 2 of 10 CFR 50.54(m)(2)(i), this section of the regulation considers a unit to be operating when it is in a mode (as defined by the unit's technical specifications) other than cold shutdown or refueling. Conversely, a nuclear power unit would be considered to be not operating when it is in a mode of cold shutdown or refueling. However, a permanently defueled facility is not in any mode. Therefore, the applicability of this requirement to a permanently defueled facility appears to lack clarity.

The purpose of 10 CFR 55 is to establish the procedures and criteria for the issuance of licenses to operators and senor operators of licensed facilities. The license requirements of 10 CFR 55.3 state that "A person must be authorized by a license issued by the Commission to perform the function of an operator or a senior operator as defined in this part." An operator is defined as "any individual licensed under this part to manipulate a control of a facility." This part also defines controls as "apparatus and mechanisms the manipulation of which directly affects the reactivity or power level of the reactor." Because the SONGS license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel, there are no longer any "controls," as defined in 10 CFR 55.4, at SONGS.

6

Based on the scope of 10 CFR 55, "Operators' Licenses," which requires licensed individuals only for the manipulation of the controls of a utilization facility, SONGS no longer requires operators licensed under this part. SCE's understanding was that 10 CFR 50.54(m) was consequently rendered moot, since this regulation serves to specify staffing levels for operators licensed pursuant to 10 CFR 55 (in that 10 CFR 55 no longer requires licensed operators at SONGS).

In summary, upon submittal of the 10 CFR 50.82 defueled certifications, the facility is no longer authorized to operate the facility or emplace or retain fuel in the reactor. As such, it is no longer possible for any "apparatus and mechanisms" to "directly affect the reactivity or power level of the reactor." Consequently, there is no longer a requirement for operators licensed under 10 CFR 55; and, per § 55.2, this regulation (10 CFR 55) no longer need apply. With no requirement for licensed operators, it follows that there would be no requirement for minimum licensed operator staffing per 10 CFR 50.54(m).

Recently however, NRC staff verbally informed SCE of their interpretation that 10 CFR 50.54(m) continues to apply to facilities that have submitted a certification of permanent removal of fuel from the reactor vessel. To address NRC concerns and to provide for shift staffing qualifications appropriate to the permanently defueled condition such that decommissioning funding is properly preserved, exemption from 10 CFR 50.54(m) and 10 CFR 55 are requested.

Summary The underlying purpose of 10 CFR 50.54(m) and 10 CFR 55 is to ensure that all operating nuclear power units are adequately staffed with licensed personnel.

The scope and radiological consequences of accidents possible at SONGS are substantially lower than those at an operating plant. Because of the significantly reduced scope and consequences of radiological events still possible at the site, the scope of operator actions and corresponding requirements for licensed operators may be accordingly reduced. Thus, the underlying purpose of the regulations will not be adversely affected by eliminating the requirements for licensed operators to fulfill on-site staffing requirements. All necessary activities can be appropriately fulfilled by certified fuel handlers and non-licensed operators.

SONGS is no longer an operating nuclear power unit. There are no design basis accidents or other credible events for SONGS that would result in a radiological dose beyond the exclusion area boundary that would exceed the Environmental Protection Agency's (EPA) Protective Action Guidelines (PAGs). Therefore, application of all of the standards and requirements in 10 CFR 50.54(m) and 10 CFR 55 are not necessary to achieve the underlying purpose of those rules.

7

I1l.

JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 10 CFR 50.12 states that the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of 10 CFR 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security.

10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present. As discussed below, these exemption requests satisfy the provisions of Section 50.12.

A. The exemptions are authorized by law The proposed exemptions to the requirements of 10 CFR 50.54(m) and 10 CFR 55 would allow Southern California Edison (SCE) to replace reliance on operators licensed pursuant to 10 CFR 55 at San Onofre Nuclear Generating Station (SONGS), with certified fuel handlers and non-licensed operators, to comport to the permanently defueled condition of the station. As stated above, 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR 50. The proposed exemptions would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. Therefore these exemptions are authorized by law.

B. The exemptions will not present an undue risk to public health and safety The underlying purpose of 10 CFR 50.54(m) and 10 CFR 55 is to ensure that all operating nuclear power units are adequately staffed with licensed personnel. As discussed within this request, it is no longer possible for the radiological consequences of design basis accidents or other credible events at SONGS to exceed the limits of the Environmental Protection Agency (EPA) Protective Actions Guidelines (PAGs) at the exclusion area boundary (EAB). Therefore, there is no longer a need for operators and senior operators licensed pursuant to 10 CFR 55 on shift at all times to respond to normal and emergency conditions. Based on the significantly reduced consequences of radiological events still possible at the site in a permanently defueled condition, such actions can be adequately addressed by an equivalent number of certified fuel handlers and non-licensed operators. Shift staffing levels are not being reduced from that previously required by 10 CFR 50.54(m), only the requirement that these onshift staff be licensed pursuant to 10 CFR 55 is being eliminated. Because appropriate training requirements for certified fuel handlers meets ANSI N18.1 and 10 CFR 50.120 and was submitted to the NRC for review and approval, and shift staffing levels will be controlled via the plant Technical Specifications, eliminating the requirement that onshift operators be licensed pursuant to 10 CFR 55 will not adversely affect SCE's ability to physically secure the site or protect special nuclear material. Therefore, the underlying purpose of the regulations will continue to be met. Since the underlying purpose of the rules 8

will continue to be met, the exemptions will not present an undue risk to the public health and safety.

C. The exemptions are consistent with the common defense and security The proposed exemptions would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS, with certified fuel handlers and non-licensed operators, to comport to the permanently defueled condition of the station.

The significantly reduced consequences of radiological events still possible at a permanently defueled facility allows for replacing reliance on operators licensed pursuant to 10 CFR 55 with an equivalent number of certified fuel handlers and non-licensed operators. Because appropriate training requirements for certified fuel handlers meets ANSI N18.1 and 10 CFR 50.120 and was submitted to the NRC for review and approval, and shift staffing levels will be controlled via the plant Technical Specifications, eliminating the requirement that onshift operators be licensed pursuant to 10 CFR 55 will not adversely affect SCE's ability to physically secure the site or protect special nuclear material. Physical security measures at SONGS are not affected by the requested exemptions. Therefore, the proposed exemptions are consistent with the common defense and security.

D. Special Circumstances Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to its regulations unless special circumstances are present. SCE believes that special circumstances are present as discussed below.

Special circumstances exist at SONGS because the plant is permanently shutdown and defueled and the radiological source term at the site is reduced from that associated with, reactor power operation. With the reactor power plant permanently shutdown and defueled, the design basis accidents and transients postulated to occur during reactor operation are no longer possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation no longer exists (Reference 7). Additionally, due to the radioactive decay of short lived isotopes, there is a continuing reduction in the potential radiological source term following the SONGS Unit 2 shutdown on January 10, 2012, and SONGS Unit 3 shutdown on January 31, 2012.

1. Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. (10 CFR 50.12(a)(2)(ii))

The underlying purpose of 10 CFR 50.54(m) and 10 CFR 55 is to ensure that all operating nuclear power units are adequately staffed with licensed personnel.

9

The radiological consequences of accidents possible at SONGS are substantially lower than those at an operating plant. Because of the significantly reduced consequences of radiological events still possible at the site, the scope of operator actions and corresponding requirements for operator staffing levels may be accordingly reduced. Thus, the underlying purpose of the regulations will not be adversely affected by eliminating the requirements for licensed operators to fulfill on-site staffing requirements. All necessary activities can be appropriately fulfilled by certified fuel handlers and non-licensed operators.

There are no design basis accidents or other credible events for SONGS that would result in a radiological dose beyond the exclusion area boundary that would exceed the Environmental Protection Agency's (EPA) Protective Action Guidelines (Reference 7). Therefore, application of the requirements in 10 CFR 50.54(m) and 10 CFR 55 are not necessary to achieve the underlying purpose of this rule.

Since the underlying purpose of the rule would be achieved by allowing SCE to replace the licensed operator staffing level requirement of 10 CFR 50.54(m) with an equivalent number of certified fuel handlers and non-licensed operators, to appropriately reflect operation of the permanently defueled facility, the special circumstances required by 10 CFR 50.12(a)(2)(ii) exist.

2. Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. (10 CFR 50.12(a)(2)(iii))

After the termination of reactor operations at SONGS and the permanent removal of the fuel from the reactor vessel, there are no design basis accidents or other credible events for SONGS that would result in a radiological dose beyond the exclusion area boundary that would exceed the Environmental Protection Agency's (EPA) Protective Action Guidelines (PAGs). Therefore, application of the requirements in 10 CFR 50.54(m) and 10 CFR 55 would result in undue costs being incurred for the maintenance of a licensed operator training and qualification program, which would be needed to maintain operators licensed pursuant to 10 CFR 55, to respond to the diminished scope of credible events.

Other licensees similarly situated, such as Zion and Millstone Unit 1, do not require operators licensed pursuant to 10 CFR 55 for shift staffing. These additional costs would result in an unnecessary adverse effect on decommissioning funding.

Therefore, since compliance with the rule would result in an undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by 10

others similarly situated, the special circumstances required by 10 CFR 50.12(a)(2)(iii) exist.

3. The exemption would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption. (10 CFR 50.12(a)(2)(iv))

The plant is permanently shutdown and defueled and the radiological source term at the site is reduced from that associated with reactor power operation.

With the reactor power plant permanently shutdown and defueled, the design basis accidents and transients postulated to occur during reactor operation are no longer possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation no longer exists. Additionally, due to the radioactive decay of short lived isotopes, there is a continuing reduction in the potential radiological source term following the following the SONGS Unit 2 shutdown on January 10, 2012, and SONGS Unit 3 shutdown on January 31, 2012.

The proposed exemptions to the requirements of 10 CFR 50.54(m) and 10 CFR 55 would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS, with certified fuel handlers and non-licensed operators, to comport to the permanently defueled condition of the station. There is no longer a need for operators and senior operators licensed pursuant to 10 CFR 55 on shift at all times to respond to normal and emergency conditions. Based on the significantly reduced consequences of radiological events still possible at the site in a permanently defueled condition, such actions can be adequately addressed by an equivalent number of certified fuel handlers and non-licensed operators.

Shift staffing levels are not being reduced from that previously required. by 10 CFR 50.54(m), only the requirement that these onshift staff be licensed pursuant to 10 CFR 55 is being eliminated. Because appropriate training requirements for certified fuel handlers meets ANSI N18.1 and 10 CFR 50.120 and was submitted to the NRC for review and approval, and shift staffing levels will be controlled via the plant Technical Specifications, eliminating the requirement that onshift operators be licensed pursuant to 10 CFR 55 will not adversely affect SCE's ability to physically secure the site or protect special nuclear material.

Eliminating the 10 CFR 50.54(m) reliance on licensed operators for shift staffing levels would obviate the need and resultant expense of the associated licensed operator training and qualification program. These reduced expenses would directly result in lower unnecessary expenditures from the decommissioning fund, which would result in benefit to the public health and safety and thereby compensate for any decrease in safety.

11

Therefore, since granting the exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, the special circumstances required by 10 CFR 50.12(a)(2)(iv) exist.

IV.

ENVIRONMENTAL CONSIDERATION The proposed exemptions meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(25), because the proposed exemption involves: (i) no significant hazards consideration; (ii) no significant change in the types or significant increase in the amounts of any effluent that may be released offsite; (iii) no significant increase in individual or cumulative occupational radiation exposure; (iv) no significant construction impact; (v) no significant increase in the potential for consequences from radiological accidents; and (vi) the requirements from which the exemption is sought involve surety, insurance or indemnity requirements or other requirements of an administrative nature.

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemptions.

(i)

No significant hazards consideration Southern California Edison (SCE) has evaluated the proposed exemptions to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92 as discussed below:

1. Do the proposed exemptions involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed exemptions would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS, with certified fuel handlers and non-licensed operators, to comport to the permanently defueled condition of the station. The proposed exemptions have no effect on plant systems structures and components (SSCs) and no effect on the capability of any plant SSC to perform its design function. The proposed exemptions would not increase the likelihood of the malfunction of any plant SSC. The proposed exemptions would have no effect on any of the previously evaluated accidents in the San Onofre Nuclear generating Station (SONGS) Updated Final Safety Analysis Report. Reliance on certified fuel handlers and non-licensed operators allowed under the exemption will not affect the probability of occurrence of any previously analyzed accident.

Therefore, the proposed exemptions do not involve a significant increase in the probability or consequences of an accident previously evaluated.

12

2. Do the proposed exemptions create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed exemptions do not involve a physical alteration of the plant. No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemptions.

Similarly, the proposed exemptions would not physically change any structures, systems or components involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created.

Furthermore, the proposed exemptions do not create the possibility of a new accident as a result of new failure modes associated with any equipment or personnel failures. No changes are being made to parameters within which the plant is normally operated, or in the setpoints which initiate protective or mitigative actions, and no new failure modes are being introduced.

Therefore, the proposed exemptions do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Do the proposed exemptions involve a significant reduction in a margin of safety?

The proposed exemptions do not alter the design basis or any safety limits for the plant. The proposed exemptions do not impact station operation or any plant SSC that is relied upon for accident mitigation.

Therefore, the proposed exemptions do not involve a significant reduction in a margin of safety.

Based on the above, SCE concludes that the proposed exemptions present no significant hazards consideration, and, accordingly, a finding of "no significant hazards consideration" is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed exemptions.

There are no materials or chemicals introduced into the plant that could affect the characteristics or types of effluents released offsite. In addition, the method of operation of waste processing systems will not be affected by the exemptions. The proposed exemptions will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents.

Therefore, the proposed exemptions will result in no significant change to the types or significant increase in the amounts of any effluents that may be released offsite.

13

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The exemptions would result in no expected increases in individual or cumulative occupational radiation exposure on either the workforce or the public. There are no expected changes in normal occupational doses. Likewise, design basis accident dose is not impacted by the proposed exemptions.

(iv) There is no significant construction impact.

There are no construction activities associated with the proposed exemptions.

(v)

There is no significant increase in the potential for consequences from radiological accidents.

See the no significant hazards considerations discussion in item 1 above.

(vi) The requirements from which exemptions are sought involve surety, insurance or indemnity requirements or other requirements of an administrative nature.

The requirements from which exemptions are sought are administrative in nature.

V.

CONCLUSION Pursuant to the provisions of 10 CFR 50.12, "Specific exemptions," Southern California Edison (SCE) is requesting exemptions from 10 CFR 50.54(m) and 10 CFR 55 for San Onofre Nuclear Generating Station (SONGS). The proposed exemptions would allow SCE to replace reliance on operators licensed pursuant to 10 CFR 55 at SONGS, with certified fuel handlers and non-licensed operators, to comport to the permanently defueled condition of the station.

These requested exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security, and special circumstances are present as set forth in 10 CFR 50.12(a)(2).

REFERENCES

1. Letter from P. T. Dietrich (SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated June 28, 2013;

Subject:

Permanent Removal of Fuel from the Reactor Vessel, San Onofre Nuclear Generating Station Unit 3

2. Letter from P. T. Dietrich (SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated July 22, 2013;

Subject:

Permanent Removal of Fuel from the Reactor Vessel, 14

San Onofre Nuclear Generating Station Unit 2

3. Letter from R. St. Onge ( SCE) to the U.S. Nuclear Regulatory Commission (NRC) dated August 20, 2013;

Subject:

Request for Approval of the Safe Storage Shift Manager/Certified Fuel Handler Training Program, San Onofre Nuclear Generating Station Units 2 and 3

4. Federal Register, Nuclear Regulatory Commission, Final Rule, Licensed Operator Staffing at Nuclear Power Units; July 11, 1983 (48 FR 31611)

5. SECY-00-0145, Rulemaking Issue Notation Vote, "Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning", dated June 28, 2000

6. Memorandum from William C.

Huffman (NRC),

"Transmittal of Report on Determination of Applicability of Title 10 of the Code of Federal Regulations to Decommissioning Nuclear Power Plants", dated July 7, 2000

7. Updated Final Safety Analysis Report - "Chapter 15 Accident Analysis" 15

ENCLOSURE 2 REVISED UPDATED FINAL SAFETY ANALYSIS "CHAPTER 15 ACCIDENT ANALYSIS" SECTIONS 15.0 AND 15.7

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS

15. ACCIDENT ANALYSES 15.0 TRANSIENT ANALYSES This chapter presents analytical evaluation of the response of the plant to postulated disturbances in process variables and to postulated malfunctions of failures of equipment. These incidents are postulated and their consequences analyzed despite the many precautions which are taken in the design, construction, quality assurance, and plant operation to prevent their occurrence. The potential consequences of such occurrences are then examined to determine their effect on the plant, to determine whether plant design is adequate to minimize consequences and to assure that the health and safety of the public and plant personnel are protected from the consequences of even the most severe of the hypothetical incidents analyzed.

The structure of this section is based on the eight by three matrix specified in Reference 1.

Initiating events are placed in one of eight categories of process variable perturbation specified in Reference I and are discussed in subsection 15.0.1. The frequency of each incident(a) was estimated, and each incident was placed in one of three frequency categories specified in Reference 1 and discussed in subsection 15.0.1.

In addition, a miscellaneous events category is established in section 15.9. This category was established specifically to include the Asymmetric Steam Generator Transient (ASGT). The ASGT does not conveniently fit into any of the other categories and was not specified by Reference 1.

(a) Incidents are defined in this section as either the initiating event or initiating event in combination with one or more coincident component or system malfunction and the resulting transient.

15.0-1 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS 15.0.1 IDENTIFICATION OF CAUSES AND FREQUENCY CLASSIFICATION 15.0.1.1 Safety Analyses Applicable after Permanent Cessation of Power Operation Per References 4 and 5, SONGS has permanently ceased operation and removed all nuclear fuel from both units reactor vessels. The irradiated fuel will be stored in the spent fuel pool (SFP) and in the Independent Spent Fuel Storage Installation (ISFSI) until it is shipped offsite. In this configuration, the SFP and its systems are dedicated only to spent fuel storage. In this condition, the number of credible accidents/transients is significantly smaller than for a plant authorized to operate the reactor or emplace or retain fuel in the reactor vessel.

Accident/transients that are no longer applicable in a permanently defueled condition are labeled as HISTORICAL throughout UFSAR Chapter 15, where appropriate. Other sections of the UFSAR may still reference to historical Chapter 15 analyses; any such references should also be considered historical. With irradiated fuel being stored in the SFP and the ISFSI, the reactor, Reactor Coolant System (RCS) and secondary system are no longer in operation and have no function related to storage of irradiated fuel. With the permanent cessation of power operation and the permanent removal of the fuel from the reactor core, the accident/transient initial conditions/initial reactor power level of the reactor core cannot be achieved and, as such, most of the accident/transient scenarios are not possible. Therefore, the postulated UFSAR Chapter 15 accidents/transients involving failure or malfunction of the reactor, RCS or secondary system are no longer applicable. UFSAR Chapter 15 accidents/transients that are applicable include:

• Radioactive Waste Gas System Leak or Failure.

• Radioactive Waste System Leak or Failure (Release to Atmosphere).

" Postulated Radioactive Release Due to Liquid Tank Failures.

• Design Basis Fuel Handling Accident Inside Fuel Building.

• Spent Fuel Cask Drop Accidents.

" Spent Fuel Pool Boiling Accident.

" Use of Miscellaneous Equipment Under 2000 lbs.

The analysis of incidents considered in this chapter are presented according to the format explained by table 15.0-1 and illustrated in the Table of Contents for this section. The initiating events are each placed in one of the categories of process variable perturbations listed in table 15.0-1. The initiating events for which analyses are presented are listed in table 15.0-2 along with their respective section designations.

Certain initiating events which are suggested for consideration in reference 1 have not been explicitly analyzed. These initiating events, along with the reasons for omission of their analyses, are provided in the appropriate paragraphs in this chapter.

The frequency of each incident has been estimated and each incident has been placed in one of the frequency categories listed in table 15.0-1. These frequency categories are defined as follows:

15.0-2 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS A. Moderate Frequency Incidents These are incidents, any one of which may occur during a calendar year for a particular plant.

B. Infrequent Incidents These are incidents, any one of which may occur during the lifetime of a particular plant.

C. Limiting Faults These are incidents that are not expected to occur but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material.

Certain malfunctions such as a stuck control element assembly (CEA) and coincident loss of normal AC power and coincident iodine spiking have been analyzed without explicit consideration of their effect on the incident frequency. The extremely low probability of these occurrences, combined with the probability of the initiating event, would produce an incident probability greatly less than that of the initiating event alone.

15.0.2 SYSTEMS OPERATION During the course of any incident, various systems may be called upon to function. These systems are described in chapter 7 and include those systems designed to perform a safety function (see sections 7.2 through 7.6); i.e., the operation of which is necessary to mitigate the consequences of the incident, and those systems not required for safety (see section 7.7).

15.0-3 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.0-1 CHAPTER 15 SUBSECTION DESIGNATION Each subsection is identified as 15.W.X.Y.Z with trailing zeros omitted where:

W = I Increase in heat removal by the secondary system (turbine plant) 2 Decrease in heat removal by the secondary system (turbine plant) 3 Decrease in reactor coolant system flowrate 4

Reactivity and power distribution anomalies 5 Increase in reactor coolant inventory 6 Decrease in reactor coolant inventory 7 Radioactive release from a subsystem or component 8 Anticipated transients without scram 9 Miscellaneous X =

I Moderate frequency incidents 2

Infrequent incidents 3 Limiting faults Y =

Initiating event (see subsection 15.0.1)

Z =

I Identification of causes and frequency classification 2

Sequence of events and systems operation 3 Core and system performance 4

Barrier performance 5 Radiological consequences 15.0-4 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.0-2 CHAPTER 15 INITIATING EVENTS (Sheet I of 3)

Paragraph Event Moderate Frequency Incidents 15.1.1.1 Decrease in feedwater temperature 15.1.1.2 Increase in feedwater flow 15.1.1.3 Increased main steam flow 15.1.1.4 Inadvertent opening of a steam generator atmospheric dump valve 15.2.1.1 Loss of external load 15.2.1.2 Turbine trip 15.2.1.3 Loss of condenser vacuum 15.2.1.4 Loss of normal AC power 15.3.1.1 Partial loss of forced reactor coolant flow 15.4.1.1 Uncontrolled CEA withdrawal from a subcritical or low power condition 15.4.1.2 Uncontrolled CEA withdrawal at power 15.4.1.3 CEA misoperation 15.4.1.4 CVCS malfunction (inadvertent boron dilution) 15.4.1.5 Startup of an inactive reactor coolant system pump 15.5.1.1 CVCS malfunction 15.5.1.2 Inadvertent operation of the ECCS during power operation 15.9.1.1 Asymmetric steam generator transient 15.0-5 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.0-2 CHAPTER 15 INITIATING EVENTS (Sheet 2 of 3)

Paragraph Event Infrequent Incidents 15.1.2.1 Decrease in feedwater temperature(a) 15.1.2.2 Increase in feedwater flow(a) 15.1.2.3 Increased main steam flow(a) 15.1.2.4 Inadvertent opening of a steam generator atmospheric dump valve(')

15.2.2.1 Loss of external load(a) 15.2.2.2 Turbine trip(a) 15.2.2.3 Loss of condenser vacuum(a) 15.2.2.4 Loss of normal AC power(a) 15.2.2.5 Loss of normal feedwater flow 15.3.2.1 Total loss of forced reactor coolant flow 15.3.2.2 Partial loss of forced reactor coolant flow(a) 15.5.2.1 CVCS malfunction~a)

Limiting Faults 15.1.3.1 Steam system piping failures 15.2.3.1 Feedwater system pipe breaks 15.2.3.2 Loss of normal feedwater flow(b) 15.3.3.1 Reactor coolant pump shaft seizure 15.3.3.2 Single reactor coolant pump sheared shaft 15.3.3.3 Complete loss of forced reactor coolant flow(b) 15.4.3.1 Inadvertent loading of a fuel assembly in an improper position 15.0-6 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS Table 15.0-2 CHAPTER 15 INITIATING EVENTS (Sheet 3 of 3)

Paragraph Event 15.4.3.2 CEA ejection 15.6.3.1 Primary sample or instrument line break 15.6.3.2 Steam generator tube rupture 15.6.3.3 Loss of coolant accident 15.6.3.4 Inadvertent opening of a pressurizer safety valve 15.7.3.1 Radioactive waste gas system leak or failure 15.7.3.2 Radioactive waste system leak or failure (release to atmosphere) 15.7.3.3 Postulated radioactive releases due to liquid tank failures 15.7.3.4 Design Basis fuel handling accident inside fuel building 15.7.3.5 Spent fuel cask drop accidents 15.7.3.6 Spent fuel pool gate drop accident 15.7.3.7 Test equipment drop 15.7.3.8 Spent fuel pool boiling accident 15.7.3.9 Design basis fuel handling accident inside containment 15.7.3.10 Spent fuel assembly drop 15.7.3.11 Use of miscellaneous equipment under 2000 lbs 15.8 1_Anticipated transient without scram (ATWS)

TABLE NOTES (a) These incidents involve the same initiating event as the corresponding moderate frequency incidents but include either a concurrent single active component failure or single operator error.

(b) These incidents involve the same initiating event as the corresponding infrequent incidents but include either a concurrent single active component failure or single operator error.

The engineered safety feature system (ESF) and systems required for safe shutdown are described in section 7.3 and 7.4, respectively. The manner in which these systems function during incidents is discussed in each incident description.

15.0-7 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS The instrumentation which is required to be available to the operator to assist in evaluating the nature of the incident and in determining the required action is described in section 7.5. The use of this instrumentation by the operator during each incident is discussed in each incident description.

Systems which are not required to perform safety functions are described in section 7.7. No credit is taken in the analysis for any operator action prior to initiation of the event which could normally mitigate the consequences of the transient; however, the analyses are performed on the basis that the plant is being operated within all limiting conditions for operation at the initiation of all events.

The effects of malfunctions of single active components or system and/or operator errors are considered as noted in the discussions of specific incidents.

15.0.3 CORE AND SYSTEM PERFORMANCE 15.0.3.1 Mathematical Model The nuclear steam supply system. (NSSS) response to various incidents was simulated using digital computer programs and analytical methods, most of which are documented in reference 2 and have been approved for use by the NRC in reference 3. Most of those programs and methods not documented in reference 2 are documented in topical reports which have been submitted to the NRC for review, and are referenced herein.

15.0.3.1.1 DELETED 15.0.3.1.2 DELETED 15.0.3.1.3 DELETED 15.0.3.1.4 DELETED 15.0.3.1.5 DELETED 15.0.3.1.6 DELETED 15.0.3.1.7 DELETED 15.0.3.1.8 DELETED 15.0-8 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS 15.0.3.1.9 Reactor Physics Computer Programs Numerous computer programs are used to produce the input reactor physics parameters required by the NSSS simulation and reactor core program previously described. These reactor physics computer programs are described in chapter 4.

15.0.3.1.10 DELETED 15.0.3.1.11 DELETED 15.0.3.2 Initial Conditions The incident discussed in this section have been analyzed over a range of values for the principal process variables.

15.0.3.3 Input Parameters The ranges of input parameters considered in the analysis are reflected in chapter 15.

In some event analyses, parameter values outside the ranges specified herein, or more conservative than the values specified herein, were used. This has the effect of introducing additional conservatism into the results.

15.0.3.3.1 DELETED 15.0.3.3.2 DELETED 15.0.3.3.3 DELETED 15.0.3.3.4 DELETED 15.0.3.3.5 DELETED 15.0.3.3.6 DELETED 15.0.3.3.7 DELETED 15.0.3.3.8 DELETED 15.0.3.3.9 DELETED 15.0.4 BARRIER PERFORMANCE 15.0.4.1 Mathematical Model 15.0-9 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS The mathematical model used for evaluation of barrier performance is identical to that described in paragraph 15.0.3.1.

15.0.4.2 Initial Conditions The initial conditions used for evaluation of barrier performance are identical to those described in paragraph 15.0.3.2.

15.0.4.3 Input Parameters The input parameters used for evaluation of barrier performance are identical to those described in paragraph 15.0.3.3.

15.0.5 RADIOLOGICAL CONSEQUENCES This subsection summarizes the assumptions, parameters, and calculational methods used to determine the doses that result from postulated accidents.

San Onofre Units 2 and 3 are licensed for full implementation of the Alternative Source Term (AST) methodology for radiological consequence analyses. All radiological analyses performed to show compliance with regulatory requirements shall address all characteristics of the AST and the Total Effective Dose Equivalent (TEDE) criteria of I OCFR50.67.

Appendix 15G identifies the models used to calculate offsite and control room radiological doses due to postulated accidents evaluated in accordance with the AST dose analysis methodology of Regulatory Guide 1.183. A list of the accidents modeled using AST methodology is provided in section 15G. 1.

The definition of a limiting fault, as provided in subsection 15.0.1, is an incident that is not expected to occur but is postulated because its consequences include the potential for the release of significant amounts of radioactive materials. For the design basis case, very conservative assumptions are made regarding the event parameters. The parameters that have been modified for the realistic analyses are presented in the description of each limiting fault.

Information used repetitively throughout the section is provided in appendix 15G for AST radiological calculations. These appendices contain information on dose models, atmospheric dispersion factors, control room parameters, and activity release models.

15.0.6 DELETED 15.0.7 DELETED 15.0.8 DELETED 15.0-10 Rev: 34

San Onofre 2&3 FSAR Updated TRANSIENT ANALYSIS 15.

0.9 REFERENCES

1.

NRC Regulatory Guide 1.70, Revision-2, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," September 1975.

2.

"Combustion Engineering Standard Safety Analysis Report," CESSAR Docket No. STN 50-470, December 1975.

3.

"Combustion Engineering Standard Safety Analysis Report (CESSAR System 80 Nuclear Steam Supply System Standard Nuclear Design Preliminary Design Approval," PDA-2, Docket No. STN 50-470, NRC, December 31, 1975.

4.

Letter from Peter T. Dietrich to U.S. Nuclear Regulatory Commission, "Docket No. 50-361 Permanent Removal of Fuel from the Reactor Vessel San Onofre Nuclear Generating Station Unit 2", dated July 22, 2013.

5.

Letter from Peter T. Dietrich to U.S. Nuclear Regulatory Commission, "Docket No. 50-362 Permanent Removal of Fuel from the Reactor Vessel San Onofre Nuclear Generating Station Unit 3", dated June 28, 2013.

15.0-11 Rev: 34

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT

15.

ACCIDENT ANALYSES 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 MODERATE FREQUENCY INCIDENTS Incidents in this category are postulated as limiting faults. Moderate frequency incidents will have radiological consequences less severe than the corresponding limiting fault described below.

15.7.2 INFREQUENT INCIDENTS Incidents in this category are postulated as limiting faults. Infrequent incidents will have radiological consequences less severe than the corresponding limiting fault described below.

15.7.3 LIMITING FAULTS 15.7.3.1 Radioactive Waste Gas System Leak or Failure The evaluation of the radiological consequences for a Radioactive Waste Gas System leak or failure assumes a minimum of 17 months since the shutdown of Units 2 and 3.

This event is modeled with the Alternative Source Term (AST). Additional assumptions associated with AST modeling are provided in Appendix 15G.

15.7.3.1.1 Identification of Causes and Frequency Classification The most limiting waste gas accident is defined as an unexpected and uncontrolled release to the atmosphere of the radioactive xenon and krypton fission gases and iodines that are stored in one waste gas decay tank. The gaseous radwaste system (GRS) is described in section 11.3.

Waste gas decay tanks are constructed to ASME III Class 3 requirements and are designed for 350 psig. When the waste gas decay tank pressure reaches 330 psig, an alarm is actuated and compressor operation is terminated manually. In addition, the GRS is located in the Seismic Category I auxiliary building, and the waste gas decay tanks, compressors, surge tank, piping, and valves up to, but not including the first system isolation valve are Seismic Category II Quality Class III. This accident is considered a limiting fault, and a rupture of a waste gas decay tank is analyzed to define the worst consequences of a gaseous release that could result from any malfunction in the gaseous radwaste system.

15.7-1 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.1.2 Sequence of Events and System Performance 15.7.3.1.2.1 Design Basis Sequence of Events and System Performance It is assumed that the plant has been operating at 3560 MWt (105% of the originally licensed power level of 3390 MWt) with 1% failed fuel for an extended period sufficient to achieve equilibrium radioactive concentrations in the reactor coolant. The maximum gas activity release from either plant would occur after shutdown and coolant degasification. One reactor coolant inventory stored in a single tank would provide an upper limit for stored gas activity. This tank is assumed to rupture and all of the noble gases and iodines are assumed to be released to the atmosphere in a 2-hour period, consistent with Regulatory Guide 1.24. Tables 15.7-1 and 15.7-2 list the conservative assumptions for waste gas decay tank rupture and waste gas decay tank inventory prior to release, respectively.

No credit is taken for Control Room Isolation Signal (CRIS) or the Control Room Emergency Air Cleanup System (CREACUS). For conservatism the control room dose is calculated for an individual at the control room outside air intake location. The total effective dose equivalent (TEDE) dose at this location is conservatively greater than it would be inside the Control Room.

The activity concentration inside the control room would be smaller since only a portion of the outside cloud would enter the control room envelope via ventilation system inflow or inleakage.

15.7.3.1.2.2 DELETED 15.7.3.1.3 Core and System Performance This paragraph is not applicable for a waste gas system leak or failure.

15.7.3.1.4 Barrier Performance This paragraph is not applicable for a waste gas system leak or failure.

15.7.3.1.5 Radiological Consequences 15.7.3.1.5.1 Design Basis Assumptions and Calculational Model Assumptions and methods used in this analysis are consistent with those of Regulatory Guide 1.24, except as discussed in appendix 3A.

The X/Q value (5% level) used is representative of the meteorology for the 0- to 2-hour interval at the location of the dose point; i.e., at the actual site boundary and at the outer boundary of the LPZ. The leak rate from the auxiliary building is such that the total leakage is equal to the total release of activity from the tank. The Alternative Source Term (AST) models used to calculate doses are discussed in appendix 15G.

15.7-2 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.1.5.2 DELETED Table 15.7-1 ASSUMPTIONS FOR WASTE GAS DECAY TANK RELEASE ACCIDENT Design Basis Assumption Source Data Power level prior to accident is 3,560 MWt.

RCS radioactive concentrations are maximum values based on 1%

failed fuel.

All gases stripped from processing a RCS volume are immediately passed to the gas decay tank which fails. Gas stripper partition factor is 1 for noble gases, 10-3 for iodines. Tank activity is presented in table 15.7-2.

A decontamination factor of 10 is assumed for the CVCS purification ion-exchanger for iodine.

Accident occurs immediately following a cold shutdown in both units.

No credit taken for radioactive decay during transit.

Number of holdup components and capacity as shown in table 11.3-1.

Source Data Gas surge tank, compressors, waste gas decay tanks, and interconnecting piping up to and including first normally closed isolation valves are Seismic Category II.

For primary coolant volume refer to section 11.1.

Activity Release All gases released from tank leak from auxiliary building at ground level within a 2-hour period.

Meteorological 5% level x/Qs per Appendix 15G Section 15G.3.

Data Dose Data Doses calculated using the model discussed in appendix 15G.

15.7-3 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-2 RADIOLOGICAL RELEASES AS A RESULT OF A WASTE GAS DECAY TANK RELEASE ACCIDENT Activity Release to Atmosphere (Ci)

Design Basis Decay For Isotopes Assumptions 17 Months 1-131 1.178.x 10l 9.4615 x 102 1 1-132 3.329 x 10-2 0.0 1-133 1.483 x 10-'

0.0 1-134 1.445 x 10-2 0.0 1-135 6.505 x 10-2 0.0 Kr-85m 5.925 x 102 0.0 Kr-85 1.294 x 103 1.1823 x 103 Kr-87 3.178 x 102 0.0 Kr-88 1.029 x 103 0.0 Xe-131m 6.030 x 102 0.0 Xe-133 8.425 x 10 4 0.0 Xe-135 2.331 x 103 0.0 Xe-135m 2.774 x 102 0.0 Xe-138 1.408 x 102 0.0 15.7-4 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.1.5.3 Results and Conclusions The results of a postulated waste gas decay tank rupture are presented in table 15.7-2A. In the unlikely event of rupture of a waste gas decay tank resulting in a release of stored gaseous activity from the reactor coolant system (RCS), the doses at the exclusion area boundary (EAB) and low population zone (LPZ) are less than the 100 mRem TEDE offsite dose criterion per Regulatory Issue Summary 2006-04. The dose at the Control Room is less than the 5 Rem TEDE criterion per 10 CFR 50.67.

Table 15.7-2A RADIOLOGICAL EXPOSURES AS A RESULT OF A WASTE GAS DECAY TANK RELEASE ACCIDENT DS ACCEPTANCE DOSE RECEPTOR DE CRITERION (mRem TEDE)

(mRem TEDE)

Control Room (30-day accident duration) 520 5,000 EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 0.14 100 LPZ (30-day accident duration) 0.00 100 15.7.3.2 Radioactive Waste System Leak or Failure (Release to Atmosphere)

The evaluation of the radiological consequences for a liquid Radioactive Waste System leak or failure (with release to atmosphere) does not assume any post-shutdown decay time. The doses would be less if a minimum decay time of 17 months since the shutdown of Units 2 and 3 was assumed.

This event is modeled with the Alternative Source Term (AST). Additional assumptions associated with AST modeling are provided in Appendix 15G.

15.7.3.2.1 Identification of Causes and Frequency Classification Liquid releases considered include rupture of radwaste tanks, refueling water storage tanks, primary ion-exchangers, and the blowdown demineralizer neutralization sump line. The most limiting of these is defined as an unexpected and uncontrolled release of the radioactive liquid stored in a radwaste secondary tank. The radwaste secondary tanks are Seismic Category II, Quality Class III tanks at atmospheric pressure. Rupture of these tanks is considered a limiting fault. A radwaste secondary tank rupture would release the liquid contents in the auxiliary building (radwaste area).

The radiological consequences of the release to the atmosphere of radioactive iodine and fission gases are considered in this evaluation.

15.7-5 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.2.2 Sequence of Events and System Operation 15.7.3.2.2.1 Design Basis Sequence of Events and System Operation It is assumed that radwaste secondary tank activity is based on 1% failed fuel. The basis of this maximum radwaste secondary tank activity is presented in paragraph 11.2.2.1.2. Design basis assumptions are presented in table 15.7-3. Source terms are shown in table 15.7-4.

A radwaste secondary tank is assumed to rupture, releasing the contents of the tank to the auxiliary building. All of the radioactive fission gases and iodines are assumed to be released to the outside atmosphere in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

15.7.3.2.2.2 DELETED 15.7.3.2.3 Core and System Performance This paragraph is not applicable for a radioactive waste system leak or failure.

15.7.3.2.4 Barrier Performance This paragraph is not applicable for a radioactive waste system leak or failure.

15.7.3.2.5 Radiological Consequences The assumptions used to evaluate the rupture of a radwaste secondary tank are listed in table 15.7-3 and the radioactive inventory in the tanks is listed in table 15.7-4. The Alternative Source Term (AST) models used to calculate doses are discussed in appendix 15G.

Offsite doses due to the rupture of a radwaste secondary tank are presented in table 15.7-4A. As shown, they are less than the 100 rnRem TEDE offsite dose criterion per Regulatory Issue Summary 2006-04.

15.7-6 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-3 ASSUMPTIONS FOR LIQUID TANK RUPTURE (RELEASE TO ATMOSPHERE)

Design Basis Assumption Source Data RCS radioactive concentrations are maximum values based on 1% failed fuel and 3560 MWt (105% of the originally licensed power level of 3390 MWt).

Tank activity basis as described in paragraph 11.2.2.1.2.

Decontamination factors for coolant radwaste system components as described in table 11.2-9.

Iodine partition factor after tank failure (1.0).

Activity Release Gases and iodines are released from auxiliary building at ground level within a 2-hour period.

Meteorological 5% level x/Qs per Appendix 15G Section 15G.3.

Data Dose Data Doses calculated using the model discussed in appendix 15G.

Table 15.7-4 RADIOLOGICAL RELEASES AS A RESULT OF LIQUID TANK RUPTURE (RELEASE TO ATMOSPHERE)

Radioactivity Released (Ci)

Design Basis Isotopes Assumptions 1-131 2.010 x 10-'

1-132 2.875 x 10-3 1-133 1.662 x 101 1-134 1.868 x 10-4 1-135 3.024 x 10.2 Kr-85m 1.570 x 10.1 Kr-85 2.33 x 100 Kr-87 8.506 x 10"'

Kr-88 1.281 x 10-'

Xe-131m 1.047 x 100 Xe-133 1.399 x 102 Xe-135m 3.171 X 10-4 Xe-135 1.501 x 100 Xe-138 1.316 x 10"4 15.7-7 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-4A RADIOLOGICAL EXPOSURES AS A RESULT OF LIQUID TANK RUPTURE (RELEASE TO ATMOSPHERE)

ACCEPTANCE DOSE RECEPTOR DE CRITERION (mRem TEDE)

(mRem TEDE)

EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 7.1 100 LPZ (30-day accident duration)

[

1.4 100 15.7.3.3 Postulated Radioactive Releases Due to Liquid Tank Failures 15.7.3.3.1 Identification of Causes and Frequency Classification Accidents involving release of radioactive liquids from tanks may involve rupture of tanks inside the containment, inside the auxiliary building, or of the refueling water or condensate storage tanks located outside. Tanks inside the containment include the reactor coolant drain tank and quench tank, which are designed to Seismic Category II criteria. The volume control tank and all liquid radwaste processing tanks, located in the Seismic Category I auxiliary building, are designed to Seismic Category I and II criteria, respectively. The Seismic Category I refueling water and Seismic Category II condensate storage tanks located in the yard area are surrounded by retention basins. The Seismic Category I Condensate Storage Tank is administratively controlled to ensure that any overflow will be within 10 CFR 20 limits. An accident involving a liquid tank failure is considered a limiting fault.

15.7.3.3.2 Sequence of Events and System Operations A hypothetical rupture of a tank inside the containment would release radioactive liquid to the containment sump where it would be collected and processed through the radioactive waste disposal system. The containment has a steel-lined interior structure; therefore, there is no pathway for leaked fluids to affect water in unrestricted areas.

Radioactive tanks in the auxiliary buildings are contained in separate, concrete-walled rooms.

These rooms are provided with water stops at the construction joints and seals wherever piping penetrates through the concrete walls to the tanks. Drain lines from the rooms are routed to the radwaste area sump. Spilled leakage would be collected in the sump and processed through the radioactive waste disposal system. Radioactive liquids released from a RWST or the Seismic Category II condensate storage tank would be contained in the concrete retention basins surrounding each tank. The Seismic Category I condensate storage tank is subject to administrative controls described in section 2.4.12 for outdoor, unprotected tanks that ensure any uncontrolled release of tank contents would be within 10 CFR 20 limits.

The liquid waste disposal system is designed to minimize or preclude discharge of plant-originated radioactive liquid wastes to the surrounding environment. Liquids are normally 15.7-8 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT processed and retained onsite for reuse, or are solidified and shipped offsite by an NRC-licensed disposal contractor after solidification. However, the system has optional capability of discharge to the circulating water system outfall within the limits of 1 OCFR20. As discussed in subsection 2.4.12, the radioactive waste discharge line is the only release path for radioactive effluent discharges into the surface water in an unrestricted area. Section 11.2 discusses the administrative controls and automatic interlocks, together with the fail-safe design of the instrumentation and control devices, which provide assurance against unauthorized or excessive releases of radioactive liquids.

15.7.3.3.3 Core and System Performance This paragraph is not applicable for a liquid tank failure.

15.7.3.3.4 Barrier Performance This paragraph is not applicable for a liquid tank failure.

15.7.3.3.5 Radiological Consequences No credible accident exists based on the scenarios showing that there would be no liquid release exceeding 10 CFR 20 limits. Therefore, no formal radiological consequence evaluation of an accident is warranted.

Refer to subsections 2.4.12 and 2.4.13 for a discussion of the effects of a postulated radioactive liquid tank failure on surface water and groundwater.

15.7.3.4 Design Basis Fuel Handling Accident Inside Fuel Building 15.7.3.4.1 Identification of Causes and Frequency Classification The possibility of a fuel handling accident is remote because of the many administrative controls and physical limitations imposed on the fuel handling operations (refer to subsection 9.1.4). All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor technically trained in nuclear safety and fuel handling.

Design of the fuel storage racks and handling facilities in the fuel storage area is such that fuel will always be in a subcritical geometrical array, assuming zero boron concentration in the fuel pool water. Refer to subsection 9.1.3 for a discussion of spent fuel racks. The spent fuel pool and refueling pool water contain boron at the refueling water boron concentration. Natural convection of the surrounding water provides adequate cooling of fuel during handling and storage. Adequate cooling of the water is provided by forced circulation in the spent fuel pool cooling system. At no time during the transfer from the reactor core to the spent fuel storage rack is a fuel assembly removed from the water. Fuel failure during refueling, as a result of inadvertent criticality or overheating, is not possible.

15.7-9 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT In accordance with the direction given in Sections 15.0 & 15.0.7, additional information which completes the presentation of this event is provided in Section 15.10.7.3.4.

In the fuel building, a fuel assembly could be dropped in the fuel transfer canal or in the spent fuel pool. In addition to the area radiation monitor located in the spent fuel cask area, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations. Doors in the fuel building are closed to maintain controlled leakage characteristics in the spent fuel pool region during refueling operations involving irradiated fuel.

Should a fuel assembly be dropped in the fuel transfer canal or in the spent fuel pool and release radioactivity above a prescribed level, the airborne radiation monitors sound an alarm, alerting personnel to the problem. Airborne radiation monitors in the exhaust ducts from the fuel handling building isolate the normal fuel handling building ventilation system and automatically initiate the recirculation and filtration systems. HVAC Plan Drawings 40341 and 40342 show the location of the isokinetic probes in the exhaust ducts and the general arrangement of the exhaust ducts in the fuel handling building. Interlocks and mechanical stops prevent the spent fuel cask handling crane from moving the cask over stored irradiated fuel and limit cask movement (refer to subsection 9.1.4). The probability of a fuel handling accident is very low because of the safety features, administrative controls, and design characteristics of the facility as previously mentioned. However, since the fuel handling accident is considered a limiting fault, it is postulated that a fuel assembly is dropped during refueling operations in the fuel building, breaching the cladding of the fuel pins and releasing the volatile fission products contained in the gap region of the fuel pin.

15.7.3.4.2 Sequence of Events and System Operation 15.7.3.4.2.1 Design Basis Sequence of Events and System Operation A description of the refueling procedure appears in subsection 9.1.4. The earliest anticipated time at which a spent fuel assembly would be handled is 3 days after shutdown.

For the design basis accident, the failure of 472 fuel rods was evaluated. The failure of 472 fuel rods is the largest number of fuel rods that could fail from the worst postulated assembly drop as described in paragraph 15.7.3.4.2.2.

The resultant release of radioactivity, after escaping from the spent fuel pool, is exhausted from the fuel handling building during a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period.

15.7.3.4.2.2 Structural Evaluation of Fuel Assembly The analysis assumes that a fuel assembly is dropped during fuel handling. Interlocks and procedural and administrative controls make such an event highly unlikely; however, if an assembly were damaged to the extent that one or more fuel rods were broken, the accumulated fission gases and iodines in the fuel rod gaps would be released to the surrounding water.

15.7-10 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Release of the solid fission products in the fuel would be negligible because of the low fuel temperature during refueling.

The fuel assemblies are stored within the spent fuel rack at the bottom of the spent fuel pool.

The top of the rack extends 13.2 inches above the tops of the stored fuel assemblies. A dropped fuel assembly could not strike more than one fuel assembly in the storage rack. Impact could occur only between the ends of the involved fuel assemblies, the bottom end fitting of the dropped fuel assembly impacting against the top end fitting of the stored fuel assembly. The maximum drop distance for this event is 74 inches from the bottom of a fuel assembly residing in the spent fuel handling machine to the top of a fuel assembly in the spent fuel storage racks. For the 74-inch drop, the fuel assembly impact velocity is 215 in./s and the impact stress in the fuel rod cladding is 20,100 psi. Criticality is not a concern for this postulated event, since the rack configuration remains intact.

Two cases were considered for the accidental drop of a fuel assembly onto or into the racks.

These were:

A.

Westinghouse 14 x 14 standard fuel assembly with control rods, total dry weight of 1260 pounds, dropped from a conservative height of 24.9 feet above the pool floor.

B.

Combustion Engineering 16 x 16 fuel assembly with control rods, total dry weight of 1540 pounds, dropped from a conservative height of 21.17 feet above the pool floor.

The drop orientations considered were a drop of an assembly onto the top of the racks with the assembly in a vertical position, drop of an assembly onto the top of the racks with the assembly in an inclined position, and a drop of a fuel assembly through an empty cell to the bottom of the pool.

The results of these analyses show that with 1800 ppm boron in the fuel pool water, fuel criticality does not occur. Thus, the acceptance criterion of no fuel criticality is met for all credible fuel drop accidents. Further, each of these three drop orientations was evaluated to determine the velocity of impact with the pool liner. In each case the structure at the lower end of the assembly had enough strain energy capacity to absorb the drop kinetic energy. When consideration was given to the "footprint" of the dropped assembly, the stresses imposed on the pool liner were determined not to perforate the pool liner for any of the drop accidents.

The maximum possible drop distance for a fuel assembly in the spent fuel pool is 254 inches.

This is the distance from the bottom of a fuel assembly in the spent fuel handling machine to the spent fuel pool floor. For this worst case drop, the velocity of the fuel assembly at impact with the fuel pool floor is 362 in./s and the impact stress in the fuel rod cladding is 34,000 psi.

15.7-11 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT The analyses of the fuel assembly vertical drops reported above were performed with a calculational model that incorporates skin friction and form drag of the fuel assembly into a mathematical formulation of the fuel assembly motion which is given below:

+ [(Fs + FD)/M] *2

-W/M=0 (1) where:

Fs

= skin friction coefficient FD

= form drag coefficient M

= mass of a fuel assembly W

= net weight of a fuel assembly x

= net velocity j

= acceleration The equation employed in calculating the impact stresses in the fuel rod clad is as follows:

0"1

= X, EJ (2) where:

oY I

= impact stress XK

= impact velocity E

= modulus of elasticity p

= mass density The allowable stress in the fuel rod cladding, a yield is 49,000 psi. This is the minimum yield stress value for unirradiated Zircaloy-4 and is conservative for irradiated fuel. Thus, for the worst case fuel assembly vertical drop, the impact stresses which result from absorbing the kinetic energy of the drop are below the yield stress of the clad and no fuel rod failures will occur.

The original design basis structural analysis postulated that the worst case fuel assembly horizontal impact results from a vertical drop of the maximum possible distance (254 inches) to the fuel pool floor, followed by rotation of the fuel assembly to the horizontal position. During this rotation, it is postulated that the assembly strikes a protruding structure. The fuel storage 15.7-12 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT pool is designed without such protruding structures and hence the shape and nature of the assumed member is indeterminate. For this analysis, therefore, a line load has been assumed for the most severe accident.

The original design basis structural analysis of this fuel assembly drop has revealed that the most severe impact location is between the top two spacer grids due to the relatively higher impact velocity of the top of the fuel assembly. Since the impact area is within the fuel rod upper plenum region, the fuel pellets do not provide clad support and do not enter into the failure analysis. To obtain an estimate of the number of fuel rods which might fail, the fuel assembly was modeled and calculations performed with the SHOCK(2) computer code. The SHOCK code allows modeling of the fuel assembly to include consideration of localized deformations about the impact point as well as general bending of the fuel assembly. The code's input data describing fuel material properties and pool conditions were kept consistent with the circumstances of the accident; i.e., irradiated fuel assembly material properties, water and fuel rod cladding temperatures corresponding to spent fuel pool conditions. For this worst case fuel assembly drop accident, no more than four rows of fuel rods (60 rods) would fail due to the combined bending and localized deformation which results from absorbing the kinetic energy at impact. For conservatism, fuel rod cladding failure was assumed to occur if the stress distribution across the fuel rod tube reached a uniform value equal to the yield stress of irradiated Zircaloy. The use of irradiated fuel rod properties for the horizontal impact is conservative because of the greater energy absorbing capability of unirradiated Zircaloy.

The current structural evaluation was originated to determine the extent of fuel rod damage produced by a fuel assembly (i.e., fuel bundle) being dropped from the fuel handling device and impacting one or more fuel bundles in the spent fuel rack during fuel handling operations. The structural evaluation addresses increases in the fuel bundle weight and to include the weights of components, handling grapples, and discretionary margin. Fuel rod damage is limited to 236 rods per bundle (two bundles are considered, dropped and impacted) regardless of the type(s) or number(s) of impact.

In the current structural evaluation, energy balance theory is employed to determine the number of damaged fuel rods resulting from the postulated events. The methodology used in the current structural evaluation is in keeping with the original structural analysis.

Due to the design of the spent fuel rack, each spent fuel bundle is placed in a separate rack with very small gaps between the full assembly and the rack. Moreover, the height of the rack exceeds the height of the spent fuel bundle. Therefore, the spent fuel bundle may be impacted only by a vertically falling fuel assembly hitting the bundle symmetrically, i.e. the axis of the dropped fuel assembly must coincide with the axis of the impacted bundle (asymmetrical contact is practically non-achievable, and a horizontally dropped assembly cannot hit a fuel bundle). No more than one impacted fuel bundle may be affected, and no tipping of the impacted fuel bundle is achievable. Therefore, the only loading on an impacted fuel bundle in the spent fuel rack is the result of a symmetrical impact with the vertically dropped fuel bundle. In addition, the structural 15.7-13 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT evaluation considers that the vertically dropped fuel bundle could tip over after impact with the spent fuel rack and come to rest with a horizontal impact.

For the fuel handling accident inside the fuel handling building, the current structural evaluation for the bundle drop scenario at the spent fuel rack location determines that a maximum of 472 fuel rods will fail as a result of a vertical drop of the fuel assembly for a dropped weight up to 2065 pounds. The drop weight of 2065 pounds represents a bundle dry weight of 1495 pounds, plus 120 pounds of components (e.g., control element assembly [CEA], neutron sources, etc.) plus 400 pounds of grapples, plus 50 pounds discretionary margin.

15.7.3.4.3 Core and System Performance This paragraph is not applicable for a fuel handling accident.

15.7.3.4.4 Barrier Performance This paragraph is not applicable for a fuel handling accident.

15.7.3.4.5 Radiological Consequences This section presents the assumptions, design input, methodology, and radiological consequences of a fuel handling accident inside the fuel handling building (FHA-FHB), based on the alternative source term (AST) guidance of Regulatory Guide (RG) 1.183.

Regulatory Guide 1.183 Appendix B provides assumptions for use in evaluating the radiological consequences of an FHA-FHB using AST methodology. These assumptions supplement the guidance provided in the main body of RG 1.183.

The characteristics of the FHA-FHB model are summarized in table 15.7-5. A supplemental description of the FHA-FHB model source term, and control room and offsite dose receptors, is presented in appendix 15G.

15.7-14 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-5 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT IN THE FUEL HANDLING BUILDING FHA-FHB PARAMETER MODELED VALUE Dose acceptance criteria, Rem TEDE Control Room 5

EAB 6.3 LPZ 6.3 FHA-FHB source term Maximum decay time after reactor shutdown, hours 12,240 (17 months)

Average fuel rod isotope inventory at 12,240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, curies/rod per Appendix 15G Table 15G-4 Radial peaking factor applied to all failed fuel rods 1.75 Number of failed fuel rods 472 Core fission product fractions in fuel rod gaps Iodine-131 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Alkali Metals (Cesium, Rubidium) 0.12 Fraction of gap activity released to the fuel storage pool 1.00 Minimum water depth above damaged fuel rods, feet 23 Fuel storage pool decontamination factors lodines (effective DF) 200 Noble Gases I

Particulates Infinite Iodine composition above the fuel storage pool, percent of iodine Elemental iodine 57 Organic iodide 43 Fuel Handling Building model ESFAS - fuel handling [building] isolation signal (FHIS) not modeled Post-Accident Cleanup Units (PACUs) not modeled Activity release duration from FHB, hours 2

FHB net free volume, cubic feet 365,305 FHB air exhaust flow rate, ft3/minute 22,000 Offsite dose evaluation model per Appendix 15G Section 15G.3 Control Room dose evaluation model per Appendix 15G Section 15G.4 FHA-FHB Release Points to Control Room per Section 2.3.4.2.2 and Atmospheric Dispersion Factors, seconds/mi3 Figure 6.4-3 15.7-15 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.4.5.1 FHA-FHB Source Term The fuel handling accident inside the fuel handling building involves the inadvertent dropping of a fuel assembly during fuel handling operations, and the consequent rupture of fuel pins in the dropped assembly. Consistent with RG 1.183 Appendix B Section 1.1, the number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case. Section 15.7.3.4.2.2 presents a structural evaluation which determined that a maximum of 472 fuel rods will fail as a result of the drop of a fuel assembly on to the fuel assemblies stored in fuel storage pool fuel racks.

Table 15G-4 presents the fission product inventory of an average fuel rod in the reactor core.

Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, a radial peaking factor of 1.75 is applied to the Table 15G-4 average fuel rod isotope inventory to determine the activity inventory in each of the 472 failed fuel rods.

Consistent with RG 1.183 Appendix B Section 3.1, the FHA-FHB dose analysis models 12,240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of radioactive decay prior to the event, which is also consistent with the minimum decay time required by SONGS administrative controls prior to movement of irradiated fuel in the reactor vessel.

Consistent with RG 1.183 Appendix B Section 1.2, the fission product release from the breached fuel is based on RG 1.183 Regulatory Position 3.2. Consistent with RG 1.183 Footnote 11, the release fractions are acceptable for use since the fuel has a peak burnup of less than 62,000 MWD/MTU, and a maximum linear heat generation rate that does not exceed 6.3 kw/ft peak rod average power for bumups exceeding 54 GWD/MTU.

All gap activity in the damaged rods is instantaneously released into the fuel storage pool.

Radionuclides that are considered include isotopes of xenon, krypton, iodine, bromine, cesium, and rubidium. Cesium and rubidium are particulates that are retained in the spent fuel pool water. Therefore, these radionuclides do not contribute to the FHA doses.

Consistent with RG 1.183 Appendix B Section 1.3, the chemical form of radioiodine released from the fuel to the fuel storage pool is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the fuel storage pool water and instantaneously re-evolve as elemental iodine. Consequently, the chemical form of radioiodine in the fuel storage pool, prior to application of a decontamination factor, is 99.85 percent elemental iodine and 0.15 percent organic iodide.

Per Units 2 & 3 Technical Specification LCO 3.7.16, during movement of irradiated fuel assemblies in the fuel storage pool, the fuel storage pool water level shall be at least 23 feet over the top of the irradiated fuel assemblies seated in the storage racks. As noted in the LCO 3.7.16 Bases, there would be less than 23 feet of water above the top of a dropped single bundle lying horizontally on top of the spent fuel racks. However, as also noted in the LCO 3.7.16 Bases, 15.7-16 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT when the potential of a dropped fuel assembly exists (which is when fuel is being moved) a water level is maintained that would ensure that there would be greater than 23 feet above the fuel assembly laying on top of the racks. This increased water level is required by Units 2 & 3 Technical Specification LCO 3.9.6 when the fuel storage pool is connected to the refueling cavity and by station procedures whenever fuel is being moved.

Consistent with RG 1.183 Appendix B Section 2, the 23 foot water depth requirement allows for elemental and organic iodine decontamination factors of 500 and 1, respectively, giving an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species.

Consistent with RG 1.183 Appendix B Section 3, the retention of noble gases in the water in the fuel storage pool is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the water in the fuel storage pool (i.e., infinite decontamination factor).

15.7.3.4.5.2 FHA-FHB Activity Release Model Consistent with RG 1.183 Appendix B Section 4.1, the radioactive material that escapes from the fuel storage pool to the fuel handling building is released to the environment over a 2-hour time period (i.e., FHB closure is not modeled during the FHA-FHB event).

Consistent with the 2-hour release model requirement, the FHA-FHB AST dose analysis does not model the generation of an ESFAS fuel handling [building] isolation signal (FHIS). The FHB normal ventilation exhaust is assumed to remain operational throughout the FHA-FHB event.

The FHB air volume dilutes the gaseous activity released from the damaged fuel rods.

The FHA-FHB AST dose analysis does not model a reduction in the amount of radioactive material available for release from the FHB by the fuel handling building post-accident cleanup unit (PACU) filter system. The FHB PACU system consists of two independent, redundant trains that each consists of charcoal and HEPA filters for the removal of airborne gaseous and particulate activity following a fuel handling accident.

The release of activity to the environment within the required 2-hour time period is established by specifying a FHB air exhaust flow rate that ensures that at least 99.9 percent of the gaseous activity will be released to the environment.

Activity released during the FHA-FHB event is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Activity may be released to the environment via the FHB normal ventilation exhaust system through the main 15.7-17 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT plant vent, or as leakage through FHB penetrations (e.g., doors). Table 15.7-5A presents the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors for these release pathways as discussed in Section 2.3.4.2.2. Since one set of atmospheric dispersion factors does not consistently yield less dispersion than the others over time, a composite maximum of the two release points is utilized for assessing control room dose consequences. No credit is taken for radioactive decay of the isotopes during atmospheric dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Table 15.7-5A FHA-FHB CONTROL ROOM ATMOSPHERIC DISPERSION FACTORS FHA-FHB to CR 95th Percentile Atmospheric Dispersion Factors (seconds/m 3)

Time In FHB Main Plant Vent Modeled Time__nterval Release Point Release Point Value 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 9.48E-04 1.15E-03 1.15E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 7.61E-04 6.23E-04 7.61E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.92E-04 2.14E-04 2.14E-04 1 to 4 days 2.65E-04 2.22E-04 2.65E-04 4 to 30 days 2.43E-04 2.02E-04 2.43E-04 15.7.3.4.5.3 FHA-FHB EAB and LPZ Model Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining the total effective dose equivalent for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone. Appendix 15G Section 15G.3 addresses the applicability of this guidance to the SONGS Units 2 & 3 AST FHA-FHB dose analysis as it relates to the offsite dose exposure parameters.

As discussed in Appendix 15G Section 15G.3, the FHA-FHB dose analysis considers the dose consequences of inhalation and immersion. Radioactive material in the fuel handling building is assumed to be a negligible radiation shine source to the offsite dose receptors relative to the dose associated with immersion in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the FHA-FHB event radiological criterion for the EAB and for the outer boundary of the LPZ is 6.3 Rem TEDE.

15.7.3.4.5.4 FHA-FHB Control Room Model Regulatory Guide 1.183 Regulatory Position 4.2 provides guidance to be used in determining the total effective dose equivalent for persons located in the control room. Appendix 15G Section 15.7-18 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15G.4 addresses the applicability of this guidance to the SONGS Units 2 & 3 AST FHA-FHB dose analysis as it relates to the control room dose exposure parameters.

The control room emergency air cleanup system (CREACUS) Emergency mode of operation can be actuated either manually or automatically following a Control Room Isolation Signal (CRIS).

The CRIS may be generated automatically by a Safety Injection Actuation Signal (SIAS) or by the detection of high radioactivity concentrations in the control room outside air inflow.

The Control Room (CR) dose during a design basis FHA-FHB following permanent shutdown of SONGS Units 2 and 3 is based on:

(a)

No credit for CREACUS and CRIS and no gamma radiation shine from CREACUS charcoal and HEPA filters.

(b)

CR doses are evaluated at various CR unfiltered inflow (including inleakage) flow rates. The flow rates were varied from 500 cfm to 15,000 cfm, but only the bounding CR dose is reported.

As discussed in Appendix 15G Section 15G.4, the FHA-FHB dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud. Radiation shine from contaminated air in the fuel handling building is considered negligible due to the presence of numerous intervening concrete walls and the geometric attenuation due to the distance between the FHB and the control room.

Consistent with RG 1.183 Regulatory Position 4.4, as an AST dose analysis acceptance criterion the postulated control room dose is evaluated to ensure that that it does not exceed the 5 Rem TEDE criterion established in 10 CFR 50.67.

15.7.3.4.5.5 FHA-FHB Dose Consequences The resulting FHA-FHB offsite and control room operator doses are listed in Table 15.7-6. The analysis demonstrates that the FHA-FHB event criteria are met.

Table 15.7-6 FHA-FHB DOSE CONSEQUENCES 15.7-19 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT FHA-FHB ACCEPTANCE DOSE RECEPTOR DOSE CRITERION (REM TEDE)

(REM TEDE)

Control Room (30-day accident duration) 0.06E-3 5

(0.06 mem TEDE)-

EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 0.20Em3 6.3 (0.20 mRem TEDE)

LPZ (30-day accident duration) 0.01Em3 6.3

________________________________(0.0 1 mRem TEDE)6.

15.7.3.5 Spent Fuel Cask Drop Accidents This section analyzes spent fuel cask drop events. Three situations are considered: a spent fuel cask drop into the spent fuel pool, a spent fuel cask dropped by the Cask Handling Crane onto a flat surface, and a spent fuel transfer cask drop (due to a seismic event) from the upper shelf in the cask pool back into the lower portion of the cask pool. The spent fuel transfer cask may be loaded with up to 32 fuel assemblies.

The spent fuel cask drop events are evaluated based on the ability of the cask drops to cause the release of radioactive materials. This includes consideration of the allowed travel paths of the casks, their lift heights, and the items onto which they can be dropped.

15.7.3.5.1 Cask Drop Into Spent Fuel Pool As discussed in subsection 9.1.4, the cask handling crane is prohibited from traveling over the spent fuel pool or any unprotected safety-related equipment. Thus, an accident resulting from dropping a cask or other major load into the spent fuel pool is not credible. In addition, single-failure-proof cranes will be used at Units 2 and 3 to lift a spent fuel transfer cask out of a cask pool.

15.7.3.5.2 Cask Drop to Flat Surface As discussed in subsection 9.1.4, the potential drop of a spent fuel cask is limited to less than an equivalent 30-foot drop onto a flat, essentially unyielding, horizontal surface. Thus, the radiological consequences of this accident are not evaluated. In addition, single-failure-proof cranes will be used at Units 2 and 3 to lift a spent fuel transfer cask out of a cask pool.

15.7.3.5.3 Cask Drop from Upper Shelf in the Cask Pool Even though single-failure-proof cranes will be used at Units 2 and 3 to lift a spent fuel transfer cask out of a cask pool, a drop can be postulated when the cask is placed on the upper shelf (i.e.,

step) of a cask pool for lifting yoke change-out, prior to the transfer cask being welded closed.

15.7-20 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT During this evolution, the transfer cask is not restrained and could fall back into the lower portion of the cask pool if an earthquake occurs.

It is assumed that a minimum of 17 months have elapsed since permanent discharge from the core for Unit 2 or 3 fuel assemblies that are loaded into a transfer cask. The fuel rods from all 32 fuel assemblies that may be present in a transfer cask are conservatively assumed to rupture on impact with the bottom of the cask pool. All of the radioactive iodine and noble gases present in the gap volumes of the decayed fuel rods are assumed to be released from the unwelded transfer cask.

Other than the number of fuel assemblies considered to fail, the cask drop accident is modeled identically to that of the fuel handling accident in the fuel handling building (FHA-FHB), as addressed in UFSAR Section 15.7.3.4. No ESF system is used to mitigate the Control Room, Exclusion Area Boundary (EAB) or Low Population Zone (LPZ) dose consequences of the cask drop accident event. This includes no credit for the Fuel Handling Isolation Signal (FHIS), the fuel handling building post accident cleanup unit (PACU) filtration system, the Control Room Isolation Signal (CRIS) and the control room (CR) emergency air cleanup system (CREACUS).

Doses are evaluated for various control room unfiltered intake plus unfiltered inleakage inflow rates.

The release of radioactive material to the atmosphere represents a potential exposure hazard to control room personnel and the general public at the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ). The CR, EAB and LPZ doses are calculated using the dose evaluation model described in Section 15.7.3.4 and Appendix 15G. Consistent with the FHA-FHB event, the cask drop accident radiological criteria are 5 Rem TEDE for the control room dose and 6.3 Rem TEDE for the EAB and LPZ doses. The resulting cask drop accident offsite and control room operator doses are listed in Table 15.7-6A. The analysis demonstrates that the cask drop accident criteria are met.

Table 15.7-6A CASK DROP ACCIDENT DOSE CONSEQUENCES CASK DROP ACCEPTANCE DOSE RECEPTOR DOSE CRITERION (REM TEDE)

(REM TEDE)

Control Room (30-day accident duration) 0.89E-3 5

_L_-0.89 rnRem TEDE)

EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 3.09E-3 6.3

_(3.09 mRem TEDE)

LPZ (30-day accident duration) 0.09E-3 6.3

_________________________________1 (0.09 mnRem TEDE) 63 15.7-21 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.6 Spent Fuel Pool Gate Drop Accident Table 15.7-7 DELETED The spent fuel pool gates consist of the transfer pool gate and the cask pool gate. During normal opening and closing, the gates operate on rails and are not subject to being dropped. The gates are periodically removed from the rails to perform maintenance on the gate seals. During removal and reinstallation, the gates are temporarily moved over the high density spent fuel storage racks. A gate drop accident is postulated where a gate is accidentally dropped while being carried over the racks.

Evaluation of a pool gate drop was based on a drop height of 30 inches (elevation 36' 4") above the top of the rack. This height is administratively controlled by permitting a maximum vertical clearance of 10 inches between the bottom of the gate and the bottom edge of the gate opening until the gate is laterally moved clear of the SFP. The dimensions of the gate are 41.0 x 343.5 x 0.75 inches. The weight of the gate is 4500 pounds.

The maximum penetration occurs for the case when the gate impacts the rack at 45'. The resulting penetration depth is 21.2 inches. This results in potential fuel damage in six cells.

The amount of penetration was determined from "conservation of energy"; i.e., the energy absorbed through plastic deformation of the rack was equated to the change in potential energy of the gate. Because all the deformation was assumed to occur in the rack and drag due to water was ignored, a conservative upper bound on the penetration was determined.

Energy absorbed by the rack was based on a "knife-edge" penetration of the cell wall. Because the gate thickness is only 0.75 inches, the force to initiate this type of penetration is significantly less than any other mode of penetration. The absorbed energy was calculated by conservatively assuming perfectly plastic deformation at a load which results in a shear stress equal to 57% of the minimum compressive yield strength of the cell wall. The impact location was selected such that the maximum number of fuel assemblies was affected.

These calculations were done for a Region II rack. This region is limiting because it has only one cell wall between adjacent storage locations, whereas Region I has two cell walls between adjacent storage locations.

The results of the radiological analysis show that, with no credit taken for the FHB filters, the doses at the exclusion area boundary would exceed the dose critiera of well within (less than 25% of) the 10CFR100 limits. The Control Room doses would also exceed the limits of 10CFR50 Appendix A, General Design Criterion 19.

15.7-22 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT To eliminate the potential for adverse radiological consequences, the following administrative controls provide assurance that the cask pool and transfer pool gates will not impact fuel assemblies stored in the impact zone or while reconstitution activities are in progress:

A.

The spent fuel pool gates shall not be removed from the installed position while reconstitution activities are in progress. Normal operation of opening and closing the gates is permitted.

B.

Prior to and during rigging for removal and reinstallation of the cask pool and transfer pool gates, all fuel assemblies shall be located outside the potential primary impact zones:

1.

The primary impact zone for the transfer pool gate is located within storage racks Nos.

1 and 2, which are Region 1 type racks. Cells in rows F through P and 1 through 3 are included (30 cells total).

2.

The primary impact zone for the cask pool gate is located within storage racks Nos. 7 and 8, which are Region II type racks. Cells in rows HH through SS and 51 through 54 are included (44 cells total).

15.7.3.7 Test Equipment Drop In order to assure that excessive radiological consequences do not occur due to a test equipment skid drop, an analysis was performed that assumes a drop of a 4500 lb. piece of equipment from a height of 47 feet above the pool floor. The assumed equipment consists of a 4-foot by 6-foot base with a 200-inch long vertical H-beam attached to the base at one of the 4-foot edges.

Conservative drag calculations made for this piece of equipment indicated that the equipment would impact the top of the racks with a velocity of approximately 206 in/s. The kinetic energy of the equipment is then converted into strain energy in the rack structure.

Calculations were made to determine the load required to compress a fuel rack cell. When the yield point of the cell is reached, local buckling increases rapidly as the cell is compressed. The penetration into the rack top if the equipment base is conservatively assumed to be at an angle with the horizontal of 45E when it impacts the rack was calculated. The maximum penetration in this case is approximately 16 inches.

The top of the Unit 1 fuel assembly is approximately 51.5 inches below the top of the rack.

Therefore, this drop will not result in damage to Unit 1 fuel assemblies. The top of the Units 2 and 3 fuel assemblies is approximately 13.2 inches below the top of the rack and this drop would result in damage to 14 Units 2 and 3 fuel assemblies. An additional analysis was made to determine the maximum drop height under which no fuel assembly damage results. It was determined that for a drop height of 72 inches above the top of the rack the test equipment will impact the top of the rack with a velocity of 177 in/s. The penetration into the rack top if the equipment base is conservatively assumed to be at an angle with the horizontal of 45E when it 15.7-23 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT impacts the rack was calculated. The maximum penetration in this case is 13.0 inches below the top of the rack. Given the drop penetration of 13 inches no fuel damage occurs.

Control rods stored integrally with fuel assemblies extend above the top of the fuel assembly upper end fitting about 1.4 inches for a SONGS 1 fuel assembly and 11.1 inches for a SONGS 2 and 3 fuel assembly. Control rods inserted into a SONGS 1 assembly do not increase the potential for fuel damage because a dropped test equipment skid can not penetrate the racks far enough to impact the top of the control rod. For a SONGS 2 or 3 assembly containing a CEA, analysis shows that the CEA will not be impacted during a potential drop if the test equipment skid is maintained below 11.2 inches above the tops of the racks.

These calculations were done for a Region II rack. Since this type of rack has only one cell wall between adjacent storage locations and the Region I rack has two cell walls between adjacent storage locations, the Region II rack is the limiting case.

Administrative controls will be implemented to provide assurance that the radiological consequences of these drops are acceptable. The administrative controls include the following:

A.

The height above the pool floor that the skid may be carried over rack cells which contain Unit 1 fuel assemblies shall be limited to 47 feet (elevation 64 feet 6 inches).

B.

When the skid is lowered, it shall be lowered over empty racks or rack cells containing Unit 1 fuel assemblies only.

C.

The maximum height that the skid may travel horizontally over the racks containing Unit 2 or 3 fuel assemblies without CEAs shall be 72 inches (elevation 39 feet 10 inches). A drop from this height will not damage Units 2 and 3 fuel assemblies.

D.

All Unit 2 or 3 fuel assemblies are to be removed from the test equipment skid impact zone, 10 by 12 cells, prior to lifting or lowering the skid over the high density spent fuel storage racks.

E.

The test equipment skid shall be maintained 11 inches or less above the top of the racks when passing over CEA bearing SONGS Units 2 and 3 spent fuel assemblies in the high density spent fuel storage racks.

With these controls in place, it will ensure that the fuel assemblies are not damaged, since the depth of penetration will not impact the racks at the level where the fuel assemblies are located.

Since no fuel assemblies are damaged, there are no radiological consequences for the test equipment drop.

15.7-24 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.8 Spent Fuel Pool Boiling Accident The postulated loss of all spent fuel pool (SFP) cooling is assumed to result in SFP boiling and the release of a portion of the radionuclide inventory contained in the stored spent fuel assemblies and the SFP water.

The evaluation of the radiological consequences for the SFP boiling event assumes a minimum of 17 months since the shutdown of Units 2 and 3. Appendix 15G identifies the isotopes present in spent fuel after this period of shutdown decay.

Following a loss of SFP cooling, activity releases from the spent fuel due to evaporation and boiling disperse to the Control Room, Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) locations.

The radiological consequence analysis conservatively does not differentiate between the activity release rates before and after the onset of SFP boiling. Noble gas, iodine and tritium activity present in the failed fuel rod gap spaces of fuel rods stored within the SFP is released to the SFP water at the noble gas, iodine and tritium escape rate coefficients listed in UFSAR Table 11.1-1, with the added conservatism of an assumed spiking factor of 100. The noble gas and iodine fuel rod gap fractions are consistent with Alternative Source Term (AST) methodology. The tritium fuel rod gap fraction is assumed to be the same as that for the majority of noble gas and iodine isotopes. Both before and after the onset of SFP boiling, spent fuel noble gases, iodine and tritium gas escaping from the failed fuel rod gap spaces are assumed to be instantaneously released with no hold up or iodine partitioning in the SFP water.

Tritium activity present in the SFP water prior to the loss of SFP cooling is assumed to be released at the SFP boiling rate for the duration of the event. The SFP boiling rate is conservatively greater than the SFP evaporation rate present prior to the onset of SFP boiling.

The SFP boiling rate is a function of the decay heat load, and the heat of vaporization of water.

No credit is taken for activity retention within the fuel handling building air. No credit is taken for Fuel Handling [Building] Isolation Signal (FHIS) or filtration by the Fuel Handling Building Post-Accident Cleanup Units (PACUs). All activity escaping from the SFP is assumed to be instantaneously released to the environment and atmospherically dispersed to the control room and offsite dose receptors.

No credit is taken for Control Room Isolation Signal (CRIS) or the Control Room Emergency Air Cleanup System (CREACUS). For conservatism the control room dose is calculated for an individual at the control room outside air intake location. The total effective dose equivalent (TEDE) dose at this location is conservatively greater than it would be inside the Control Room.

The activity concentration inside the control room would be smaller since only a portion of the outside cloud would enter the control room envelope via ventilation system inflow or inleakage.

15.7-25 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-7A lists the parameters used for performing the dose analysis for the postulated SFP boiling event. Additional assumptions associated with Alternative Source Term (AST) modeling are provided in Appendix 15G.

The offsite radiological doses for the postulated SFP boiling accident do not exceed 25% of the 10 CFR Part 50.67 exposure guidelines. The radiological consequences of this event are presented in Table 15.7-8.

15.7-26 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-7A PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A SPENT FUEL POOL BOILING EVENT SFP BOILING PARAMETER MODELED VALUE Number of Stored Spent Fuel Assemblies in Spent Fuel Pool 1542 (SFP)

(= total available spaces)

SFP Decay Heat Load (17 months post-shutdown), BTU/hr 3.879E6 SFP Boiling Rate, ft 3/hour 66.84 Failed Fuel, %

I Failed Fuel Escape Rate Coefficients, sec-1 Noble Gases 6.5E-8 Iodine 1.3E-8 Tritium 1.4E-11 Spiking Factor for Noble Gases, Iodine and Tritium Releases 100 from Failed Fuel Fuel Rod Gap Release Fractions Iodine-131 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Tritium 0.05 SFP Water Iodine Decontamination Factor 1

Fuel Handling [Building] Isolation Signal (FHIS) not modeled Fuel Handling Building Post-Accident Cleanup Units (PACUs) not modeled Activity Release from FHB SFP Activity Release instantaneously dispersed to dose receptor Offsite Dose Evaluation Model per Appendix 15G Section 15G.3 per Appendix 15G Section 15G.4 Control Room Dose Evaluation Model (no credit for CRIS or CREACUS operation)

FHB Release Points to Control Room Atmospheric Dispersion per Section 2.3.4.2.2 and Factors, seconds/m 3 Figure 6.4-3 15.7-27 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-8 RADIOLOGICAL CONSEQUENCES OF SPENT FUEL POOL BOILING SFP BOILING ACCEPTANCE DOSE RECEPTOR DOSE CRITERION (REM TEDE)

(REM TEDE)

Control Room (30-day accident duration) 11.96E-03 5

(11.96 mem TEDE)_______

EAB (Maximum 2-hour dose -- 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 0.08Ee03 6.3 (0.08 5emTEDE)

LPZ (30-day accident duration) 0.25Em03 6.3 (0.25 mRemn TEDE) 15.7.3.9 Design Basis Fuel Handling Accident Inside Containment The following information is HISTORICAL and describes a design basis transient or accident event that is no longer applicable in a permanently defueled condition. Refer to Section 15.0.

15.7.3.9.1 Identification of Causes and Frequency Classification The possibility of a fuel handling accident is remote because of the many administrative controls and physical limitations imposed on the fuel handling operations (refer to subsection 9.1.4). All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor technically trained in nuclear safety and fuel handling.

In accordance with the direction given in Sections 15.0 & 15.0.7, additional information which completes the presentation of this event is provided in Section 15.10.7.3.9.

Before refueling operations start, the boron concentration of the reactor is increased so that the core will be more than 5% Ap subcritical with all control element assemblies (CEAS) removed from the reactor. Prior to the removal of the reactor vessel head, verification of complete insertion of all CEAs is obtained from a visual check of the control board mounted CRT which displays the control element drive assembly (CEDM) positions. During head removal, the audible count-rate signals from the startup channels are monitored as additional assurance that CEAs are not being inadvertently removed.

Following head removal, the upper guide structure lift rig is installed and all five finger CEAs are uncoupled from their drive shafts using remote tooling. Positive indication of uncoupling is obtained by weighing each drive shaft. Satisfactory uncoupling is determined by proper weight indication. The CEDM drive shafts are removed with the upper guide structure; the CEAs remain in the core. During upper guide structure removal, the audible count rate signals from the startup channels are monitored as additional assurance that CEAs are not being inadvertently removed.

15.7-28 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT The refueling pool water contains boron at the refueling water boron concentration. Natural convection of the surrounding water provides adequate cooling of fuel during handling and storage. Adequate cooling of the spent fuel pool water is provided by forced circulation in the spent fuel pool cooling system. Adequate cooling of the reactor cavity water is provided by forced circulation in the shutdown cooling system. At no time during the transfer from the reactor core to the spent fuel storage rack is a fuel assembly removed from the water. Fuel failure during refueling, as a result of inadvertent criticality or overheating, is not possible.

During fuel handling operations, the containment is kept in an isolatable condition, with all penetrations to the outside atmosphere either closed or capable of being closed. In addition to the area and airborne radiation monitors in the containment, portable monitors capable of sounding audible alarms are located in the fuel handling area during refueling. Should a fuel assembly be dropped and release activity above a prescribed level, the radiation monitors sound an audible alarm, and personnel are evacuated.

15.7.3.9.2 Sequence of Events and System Operation 15.7.3.9.2.1 Design Basis Sequence of Events and System Operation A description of the refueling procedure appears in subsection 9.1.4. The earliest anticipated time at which a spent fuel assembly would be handled is 3 days after shutdown.

As discussed in Section 15.7.3.9.2.2, for the design basis accident, the failure of 472 fuel rods was evaluated.

The resultant release of radioactivity, after escaping from the refueling fuel pool, is exhausted from the containment during a 2-hour period.

A Control Room isolation occurs when high radiation is sensed at the Control Room outside air intake.

15.7.3.9.2.2 Structural Evaluation of Fuel Assemblies Section 15.7.3.4.2.2 documents the original design basis structural analysis methodology for a dropped fuel assembly (i.e., fuel bundle).

The current structural evaluation was originated to determine the extent of fuel rod damage produced by a fuel bundle being dropped from the fuel handling device and impacting one or more fuel bundles in the reactor core vessel during fuel handling operations. The structural evaluation addresses increases in the fuel bundle weight and to include the weights of components, handling grapples, and discretionary margin. Fuel rod damage is limited to 236 rods per bundle (two bundles are considered, dropped and impacted) regardless of the type(s) or number(s) of impact.

15.7-29 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT In the current structural evaluation, energy balance theory is employed to determine the number of damaged fuel rods resulting from the postulated events. The methodology used in the current structural evaluation is in keeping with the original structural analysis.

For the fuel handling accident inside containment, the structural evaluation examines the case of a fuel assembly (i.e., fuel bundle) being dropped from the fuel handling device and impacting one or more fuel bundles in the reactor core vessel. Fuel bundles located in the reactor core vessel may be struck by a vertically dropped fuel bundle with the impacting configuration being either symmetrical or asymmetrical. In addition, the impacted fuel bundle may be tipped over after being struck and may rotate and collide with the walls of the Core Shroud Assembly (the impacted fuel bundle in a partial core is too tall to be able to tip over and rotate to a full horizontal position). The dropped bundle can also tip over after impact with a bundle in the core.

The dropped bundle is assumed to rotate before impacting the Core Shroud Assembly wall.

The structural evaluation determines that another impact model involves the case when a dropped fuel bundle impacts several (maximum of 4, and minimum of 2) fuel bundles in the reactor core vessel. In such an event, the impact force will be distributed between the impacted fuel bundles proportionally to the area of each fuel bundle subjected to the impact. However, the total number of the subjected fuel rods will be essentially 236, since the gap between the fuel bundles in a fairly complete core is minimal. Therefore, the fuel rods will experience practically the same stress as in the case of the symmetrical vertical impact mode.

The structural evaluation also determines that for the case of a horizontally dropped bundle falling into the reactor core, the fuel bundle cannot physically fit into the core support barrel in the full horizontal position. The vertical drop and tipping scenario is sufficient to bound any damage due to possible non-vertical drops.

For the fuel handling accident inside the reactor core vessel, the current structural evaluation for the bundle drop scenario at the reactor core location determines that a maximum of 472 fuel rods will fail as a result of a vertical drop of the fuel assembly for dropped weight up to 1865 pounds.

The drop weight of 1865 pounds represents a bundle dry weight of 1495 pounds, plus 120 pounds of components (e.g., control element assembly [CEA], neutron sources, etc.) plus 200 pounds of grapples, plus 50 pounds discretionary margin.

15.7.3.9.3 Core and System Performance This paragraph is not applicable for a fuel handling accident.

15.7-30 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.9.4 Barrier Performance This paragraph is not applicable for a fuel handling accident.

15.7.3.9.5 Radiological Consequences This section presents the assumptions, design input, methodology, and radiological consequences of a fuel handling accident inside containment (FHA-IC), based on the alternative source term (AST) guidance of Regulatory Guide (RG) 1.183.

Regulatory Guide 1.183 Appendix B provides assumptions for use in evaluating the radiological consequences of an FHA-IC using AST methodology. These assumptions supplement the guidance provided in the main body of RG 1.183.

The characteristics of the FHA-IC model are summarized in table 15.7-9. A supplemental description of the FHA-IC model source term, and control room and offsite dose receptors, is presented in appendix 15G.

In accordance with the direction given in sections 15.0 and 15.0.7, additional information which completes the presentation of this event is provided in section 15.10.7.3.9.

15.7-31 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Table 15.7-9 PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT INSIDE CONTAINMENT FHA-IC PARAMETER MODELED VALUE Dose acceptance criteria, Rem TEDE Control Room 5

EAB 6.3 LPZ 6.3 FHA-IC source term Maximum decay time after reactor shutdown, hours 72 Average fuel rod isotope inventory at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, curies/rod per Appendix 15G Table 15G-4 Radial peaking factor applied to all failed fuel rods 1.75 Number of failed fuel rods 472 Core fission product fractions in fuel rod gaps Iodine-131 0.08 Krypton-85 0.10 Other noble gases (Krypton, Xenon) 0.05 Other Halogens (Iodine, Bromine) 0.05 Alkali Metals (Cesium, Rubidium) 0.12 Fraction of gap activity released to the refueling water 1.00 Minimum water depth above reactor vessel flange (and above 23 the damaged fuel rods), feet Refueling water decontamination factors lodines (effective DF) 200 Noble Gases I

Particulates Infinite Iodine composition above the refueling water, percent of iodine Elemental iodine 57 Organic iodide 43 Containment model Containment dome air circulators not modeled ESFAS - containment purge isolation signal (CPIS) not modeled ESFAS - containment isolation actuation signal (CIAS) not modeled Containment personnel airlock closure not modeled Containment equipment hatch closure not modeled Activity release duration from containment, hours 2

Containment net free volume without dome, cubic feet 1.422E+06 Containment air exhaust flow rate, ft3/minute 82,000 Offsite dose evaluation model per Appendix 15G Section 15G.3 Control Room dose evaluation model per Appendix 15G Section 15G.4 FHA-IC Release Points to Control Room per Section 2.3.4.2.2 Atmospheric Dispersion Factors, seconds/m 3 and Figure 6.4-3 15.7-32 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.3.9.5.1 FHA-IC Source Term The fuel handling accident inside containment involves the inadvertent dropping of a fuel assembly during fuel handling operations, and the consequent rupture of fuel pins in both the dropped and impacted fuel assemblies. Consistent with RG 1.183 Appendix B Section 1.1, the number of fuel rods damaged during the accident is based on a conservative analysis that considers the most limiting case. Section 15.7.3.9.2.2 presents a structural evaluation which determined that a maximum of 472 fuel rods will fail as a result of the drop of a fuel assembly on to the fuel assemblies remaining in the partially loaded core.

Table 15G-4 presents the fission product inventory of an average fuel rod in the reactor core.

Consistent with the guidance of RG 1.183 Regulatory Position 3.1, to account for differences in power level across the core, a radial peaking factor of 1.75 is applied to the Table 15G-4 average fuel rod isotope inventory to determine the activity inventory in each of the 472 failed fuel rods.

Consistent with RG 1.183 Regulatory Position 3.1, the FHA-IC dose analysis models 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of radioactive decay prior to the event, which is consistent with the minimum decay time required by SONGS administrative controls prior to movement of irradiated fuel in the reactor vessel.

Consistent with RG 1.183 Appendix B Section 1.2, the fission product release from the breached fuel is based on RG 1.183 Regulatory Position 3.2. Consistent with RG 1.183 Footnote 11, the release fractions are acceptable for use since the fuel has a peak burnup of less than 62,000 MWD/MTU, and a maximum linear heat generation rate that does not exceed 6.3 kw/ft peak rod average power for burnups exceeding 54 GWD/MTU.

All gap activity in the damaged rods is instantaneously released into the refueling water.

Radionuclides that are considered include isotopes of xenon, krypton, iodine, bromine, cesium, and rubidium. Cesium and rubidium are particulates that are retained in the refueling pool water.

Therefore, these radionuclides do not contribute to the FHA doses.

Consistent with RG 1.183 Appendix B Section 1.3, the chemical form of radioiodine released from the fuel to the refueling water is assumed to be 95% cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide. The CsI released from the fuel is assumed to completely dissociate in the refueling water and instantaneously re-evolve as elemental iodine.

Consequently, the chemical form of radioiodine in the refueling water, prior to application of a decontamination factor, is 99.85 percent elemental iodine and 0.15 percent organic iodide.

Per Units 2 & 3 Technical Specification LCO 3.9.6, during movement of irradiated fuel assemblies within containment, the refueling water level above the top of the reactor vessel flange shall be greater than or equal to 23 feet. Since the damaged fuel assemblies would be lower than the reactor vessel flange, the water depth above the damaged fuel would be greater than 23 feet. Consistent with RG 1.183 Appendix B Section 2, the 23 foot water depth requirement allows for an overall effective decontamination factor of 200 (i.e., 99.5% of the total iodine released from the damaged rods is retained by the water). This difference in 15.7-33 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT decontamination factors for elemental (99.85%) and organic iodine (0.15%) species results in the iodine above the water being composed of 57% elemental and 43% organic species.

Consistent with RG 1.183 Appendix B Section 3, the retention of noble gases in the refueling water is negligible (i.e., decontamination factor of 1). Particulate radionuclides are assumed to be retained by the refueling water (i.e., infinite decontamination factor).

15.7.3.9.5.2 FHA-IC Activity Release Model The containment personnel airlock or equipment hatch may be open during Mode 6 core alterations and movement of irradiated fuel in containment. Consistent with RG 1.183 Appendix B Section 5.3, since the containment is open during fuel handling operations, the radioactive material that escapes from the refueling water to the containment is released to the environment over a 2-hour time period (i.e., containment closure is not modeled during the FHA-IC event).

Consistent with the 2-hour release model requirement, the FHA-IC AST dose analysis does not model the generation of an ESFAS containment purge isolation signal (CPIS) or containment isolation actuation signal (CIAS). The containment purge is assumed to remain operational throughout the FHA-IC event. The containment personnel airlock and the containment equipment hatch are assumed to remain open throughout the FHA-IC event.

The containment air volume dilutes the gaseous activity released from the damaged fuel rods.

During Mode 6 refueling operations there is no SONGS Units 2 & 3 Technical Specification requirement for the containment dome air circulators or containment cooling train fans to be operable. Therefore, no credit is taken for activity dilution within the air of the containment dome space.

The FHA-IC AST dose analysis does not model a reduction in the amount of radioactive material available for release from the containment by any containment engineered safety feature (i.e.,

containment purge filters are not credited).

The release of activity to the environment within the required 2-hour time period is established by specifying a containment air exhaust flow rate that ensures that at least 99.9 percent of the gaseous activity will be released to the environment.

Activity released during the FHA-IC event is transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. Activity may be released to the environment via the containment purge system or as leakage through containment penetrations, including the containment personnel airlock or the containment equipment hatch.

Leakage from the containment personnel airlock would be exhausted via the main plant vent.

Table 15.7-10 presents the San Onofre site-specific 95th percentile meteorology atmospheric dispersion factors for these release pathways as discussed in Section 2.3.4.2.2. Since one set of atmospheric dispersion factors does not consistently yield less dispersion than the others over time, a composite maximum of the three release points is utilized for assessing control room dose consequences. No credit is taken for radioactive decay of the isotopes during atmospheric 15.7-34 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT dispersion transit to the control room or offsite dose locations. Consistent with RG 1.183 Regulatory Positions 4.1.7 and 4.2.2, no correction is made for depletion of the effluent plume by deposition on the ground.

Table 15.7-10 FHA-IC Control Room Atmospheric Dispersion Factors FHA-IC to CR 95th Percentile Atmospheric Dispersion Factors (seconds/m3)

Containment Equipment Main Plant Vent Modeled Time Interval Shell Hatch R

Release Point Release Point 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.01E-03 8.01E-04 1.15E-03 1.15E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 6.41 E-04 6.35E-04 6.23E-04 6.41E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.77E-04 1.78E-04 2.14E-04 2.14E-04 1 to 4 days 2.36E-04 2.23E-04 2.22E-04 2.36E-04 4 to 30 days 2.20E-04 2.03E-04 2.02E-04 2.20E-04 15.7.3.9.5.3 FHA-IC EAB and LPZ Model Regulatory Guide 1.183 Regulatory Position 4.1 provides guidance to be used in determining the total effective dose equivalent for persons located at or beyond the boundary of the exclusion area, including the outer boundary of the low population zone. Appendix 15G Section 15G.3 addresses the applicability of this guidance to the SONGS Units 2 & 3 AST FHA-IC dose analysis as it relates to the offsite dose exposure parameters.

As discussed in Appendix 15G Section 15G.3, the FHA-IC dose analysis considers the dose consequences of inhalation and immersion. Radioactive material in the containment is assumed to be a negligible radiation shine source to the offsite dose receptors relative to the dose associated with immersion in the radioactive plume released from the facility.

Consistent with RG 1.183 Regulatory Positions 4.1.5, 4.1.6 and 4.4 and Table 6, the FHA-IC event radiological criterion for the EAB and for the outer boundary of the LPZ is 6.3 Rem TEDE.

15.7.3.9.5.4 FHA-IC Control Room Model Regulatory Guide 1.183 Regulatory Position 4.2 provides guidance to be used in determining the total effective dose equivalent for persons located in the control room. Appendix 15G Section 15G.4 addresses the applicability of this guidance to the SONGS Units 2 & 3 AST FHA-IC dose analysis as it relates to the control room dose exposure parameters.

15.7-35 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT The control room emergency air cleanup system (CREACUS) Emergency mode of operation can be actuated either manually or automatically following a Control Room Isolation Signal (CRIS).

The CRIS may be generated automatically by a Safety Injection Actuation Signal (SIAS) or by the detection of high radioactivity concentrations in the control room outside air inflow. Per Appendix 15G Section 15G.4.2.1.1, the FHA-IC model credits CREACUS Emergency mode of operation initiation 3 minutes (180 seconds) following the start of the event, due to detection of high radioactivity concentrations in the control room outside air inflow.

As discussed in Appendix 15G Section 15G.4, the FHA-IC dose analysis considers the dose consequences of inhalation, immersion, and radiation shine from the environmental (or outside) cloud, the control room emergency HVAC filters, and the containment building.

Consistent with RG 1.183 Regulatory Position 4.4, the postulated control room dose is evaluated to ensure that that it does not exceed the AST dose analysis acceptance criterion of 5 Rem TEDE established in 10 CFR 50.67.

15.7.3.9.5.5 FHA-IC Dose Consequences The resulting FHA-IC offsite and control room operator doses are listed in Table 15.7-11. The analysis demonstrates that the FHA-IC event criteria are met.

Table 15.7-11 RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT INSIDE CONTAINMENT ACCEPTANCE FHA-IC DOSE ACPAC DOSE RECEPTOR FRAI DE CRITERION (REM TEDE)

(REM TEDE)

Control Room (30-day accident duration) 0.6 5

EAB (Maximum 2-hour dose - 0.0 to 2.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />) 1.7 6.3 LPZ (30-day accident duration)

< 0.1 6.3 15.7.3.10 Spent Fuel Assembly Drop 15.7.3.10.1 Spent Fuel Assembly Drop onto Reconstitution Station Introduction As discussed in subsection 9.1.2.2, a spent fuel assembly will be placed atop a rack spacer during fuel reconstitution. As a result, the spacer will raise the top of the spent fuel assembly above the top of the high density spent fuel storage racks. In the reconstitution station when the rack spacers are used, there is a greater potential for damage to spent fuel assemblies than in locations where the spacers are not used, if a spent fuel assembly is accidentally dropped.

15.7-36 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Summary of Methods Current procedures restrict the number of spent fuel assemblies in the reconstitution station to six. A situation could exist during fuel reconstitution where five spent fuel assemblies are placed atop rack spacers and the sixth one is being moved towards the reconstitution station and is therefore above one or all of these elevated spent fuel assemblies. The drop of a spent fuel assembly onto the reconstitution station could damage the dropped assembly, as well as some of the spent fuel rods in the assemblies located on spacers within the reconstitution station.

To prevent such an accident in the reconstitution station during fuel reconstitution, the following two administrative controls are implemented:

A. No spent fuel assembly shall be moved over any spent fuel assembly in the reconstitution station or over adjacent storage locations when spent fuel assemblies are in the reconstitution station on the rack spacers.

B. No CEA bearing spent fuel assemblies shall be placed atop rack spacers in the reconstitution station.

Additionally, a fuel assembly dropped onto the spent fuel storage racks could have the potential to topple over onto a fuel assembly located on a spacer in the reconstitution station. In this event the fuel assembly in the reconstitution station is not damaged. The damage to the dropped fuel assembly is addressed in section 15.7.3.4.

Table of Results Analyses were completed which evaluated the potential for damage to spent fuel assemblies located on reconstitution spacers in a reconstitution station. Results concluded that damage to spent fuel could occur if a spent fuel assembly is moved and subsequently dropped onto fuel located in the reconstitution station.

Conclusion The administrative controls stated above will preclude damage to spent fuel assemblies during reconstitution.

15.7.3.10.2 Spent Fuel Assembly Drop onto CEA Bearing Spent Fuel Assemblies Control Element Assemblies (CEAs) which have been replaced are stored in the spent fuel pool inserted into spent fuel assemblies. The top of a CEA comes within 2.11 inches of the top of the high density spent fuel storage rack when stored integrally with a SONGS Units 2 and 3 spent fuel assembly. The maximum potential damage occurring in the event of a spent fuel assembly 15.7-37 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT drop onto a CEA bearing spent fuel assembly, would be failure of all fuel rods in the dropped fuel assembly and all fuel rods in the impacted CEA bearing spent fuel assembly.

The radiological consequences for the failure of two fuel assemblies is addressed by the postulated fuel handling accident inside the fuel handling building in UFSAR section 15.7.3.4.

15.7.3.11 Use of Miscellaneous Equipment Under 2000 lbs Introduction Several miscellaneous pieces of equipment weighing less than 2000 lbs (e.g., equipment used for ultrasonic testing, gamma spectrometer, eddy current testing, periscope, oxide measurement devices, tools and work platforms used for fuel reconstitution, temporary underwater pumps, skimmers) are required to be located or moved over the spent fuel storage racks during refueling operations and normal spent fuel pool maintenance. Administrative controls are currently placed on the movement of loads in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over other fuel assemblies in the storage pool. Hence, this evaluation is based on the potential for damaging fuel assemblies, if it is postulated that equipment which does not exceed the weight of a fuel assembly, CEA, and associated handling equipment (i.e., less than 2000 lbs) is dropped onto other spent fuel assemblies.

Summary of Methods The restrictions on movement of loads in excess of the nominal weight of a fuel assembly, CEA, and associated handling tool over other fuel assemblies in the storage pool ensure that, if this load is dropped, this event is bounded by other load drop events. Specifically, the activity release would be less than a fuel handling accident inside the fuel handling building, which analyzes a drop weight of 2065 pounds, which represents a bundle dry weight of 1495 pounds, plus 120 pounds of components (e.g., control element assembly [CEA], neutron sources, etc.)

plus 400 pounds of grapples, plus 50 pounds discretionary margin. Moreover, any possible distortion of fuel contained in the racks would not result in a critical array.

The restrictions on movement of loads in excess of 2000 pounds, the nominal weight of a fuel assembly, CEA, and associated handling tool are administratively controlled.

Since this event is bounded by another more limiting event, there are no principal assumptions, inputs, or sequence of events to present.

Table of Results The administrative controls relative to moving loads less than 2000 lbs minimize the potential for radiological release or criticality events. However, the dose consequences of load drops on spent fuel contained in storage pool racks are bounded by the postulated fuel handling accident inside the fuel handling building in UFSAR section 15.7.3.4.

15.7-38 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT Conclusion The dose consequences of load drops on spent fuel contained in storage pool racks are bounded by the postulated fuel handling accident inside the fuel handling building in UFSAR section 15.7.3.4.

15.7-39 Rev: 30

San Onofre 2&3 FSAR Updated RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT REFERENCES

1.

Love, A. E. H., A Treatise on the Mathematical Theory of Elasticity, 4th Edition, Dover Publications, New York, New York, October 1926.

2.

Gabrielson, V. K., SHOCK - A Computer Code for Solving Lumped Mass Dynamic Systems, SCL-DR-65-35, January 1966.

15.7-40 Rev: 30

ENCLOSURE 3 Page 1 of 34

-J S0123-0-A1, Revision 41 SONGS CONDUCT OF OPERATIONS Procedure Usage Requirements Sections Information Use The user may complete the task from memory. However, the user is responsible for performing the activity in accordance with the procedure.

ALL Information Use documents that contain a specific process order are performed in the given order unless otherwise specified within the document.

QA Program Affecting 50.59 DNA 1 72.48 DNA / RX DNA Procedure Owner John Davis

SO123-0-Al S0123-0-Al Rev 41 Page 2 of 34 CONDUCT OF OPERATIONS TABLE OF CONTENTS Section PaQe 1.0 PURPOSE..........................................................................................................................................

4 2.0 SCOPE 4

3.0 RESPONSIBILITIES..........................................................................................................................

4 4.0 PRECAUTIONS AND LIMITATIONS..........................................................................................

4 5.0 PREREQUISITES I INITIAL CONDITIONS.................................................................................

4 6.0 PROCEDURE 5

6.1 Operations Management and Leadership........................................................................

5 6.1.1 Operations Management Team................................................................

5 6.1.2 Operator Fundamentals............................................................................

6 6.1.3 Nuclear Safety Culture and Conservative Decision-Making...................... 7 6.1.4 Operator Professionalism and Conduct.....................................................

8 6.1.5 Operations Leadership Role in Site Activities...........................................

9 6.1.6 Personnel Safety Culture..........................................................................

9 6.2 Control Room Operations.................................................................................................

10 6.2.1 Control Room Access..............................................................................

10 6.2.2 Control Room Board Monitoring..............................................................

10 6.2.3 Operation / Manipulation of Plant Equipment...........................................

10 6.2.4 Administrative Duties of Control Room Personnel....................................

11 6.2.5 Reactivity Management............................................................................

11 6.2.6 Control Room Communications, Briefs and Updates...............................

12 6.2.7 Control Room Command Function.........................................................

13 6.2.8 Control Room Manning............................................................................

14 6.2.9 Annunciator Response............................................................................

14 6.3 Operations Field Activities...............................................................................................

15 6.4 S h ift A c tiv itie s.......................................................................................................................

1 6 6.4.1 Operator Log Keeping..............................................................................

16 6.4.2 Operator Turnovers and Shift Relief Status Sheets..................................

17 6.4.3 Work Process.........................................................................................

18 6.5 Human Performance and Error Reduction.....................................................................

19 6.5.1 Monitoring and Assessing Performance..................................................

19 6.5.2 Evaluations of Crew Performance...........................................................

19 6.5.3 HU Events / Error Investigation.............................................................

19 6.5.4 Problem-Solving/Decision-Making/Risk Management..............................

20 6.6 Human Performance Tools............................................................................................

21 INFORMATION USE

S0123-0-Al Rev 41 S0123-0-Al Page 3 of 34 6.7 Operator Knowledge and Development.........................................................................

21 6.7.1 6.7.2 Management Standards for Training......................................................

21 Operator Training Program.....................................................................

21 6.8 Procedure Use and Adherence........................................................................................

22 6.9 A d m in istra tiv e.......................................................................................................................

2 3 6.9.1 Active License Status Requirements.......................................................

23 7.0 A CCEPTA NCE I FUNCTIO NA L CRITERIA................................................................................

24 8.0 RETENTIO N / RECO RDS...............................................................................................................

24 9.0 DEFINITIO NS..................................................................................................................................

25

10.0 REFERENCES

/ CO M M ITM ENTS............................................................................................

27 ATTACHMENT 1

2 3

4 5

N O T U S E D......................................................................................................................................

3 0 M inim um Operations Shift Crew Composition...........................................................................

31 Documented On-Shift Hours........................................................................................................

32 Entry into 10CFR50.54(x) OR 10CFR50.54(y)...........................................................................

33 Sum mary of Changes.....................................................................................................................

34 INFORMATION USE

S0123-0-Al Rev 41 S0123-0-Al Page 4 of 34 CONDUCT OF OPERATIONS 1.0 PURPOSE NOTE This procedure includes Improved Technical Specifications [ITS] information that is NOT applicable to Current Technical Specifications [CTS] and [CTS]

information that is NOT applicable in [ITS]. The [CTS] information shall be used prior to the [ITS] effective date. The [ITS] information shall be used on or after the [ITS] effective date.

1.1 Ensure all station operations activities are conducted in a professional manner that contributes to safe and reliable station operation. This procedure addresses the important elements of the Control Room environment, field operations, and activities necessary to support safe and efficient station operations as well as establishing and maintaining industry leading operating performance.

Detailed explanations, expectations and techniques to fine tune the consistent performance of these standards may be found in associated OSM guidance. This procedure provides the standards specific to the operations department and provides direction to other departments that provide support to the Operations Department in operating the plant.

2.0 SCOPE 2.1 This procedure applies to all Operations personnel and all operations activities. It also applies to personnel from other groups as they interact with Operations in providing the support necessary to achieve safe, reliable and economic operation.

3.0 RESPONSIBILITIES 3.1 Director, Operations - Responsible for establishing the standards contained in this procedure and ensuring compliance by Operations Department personnel.

3.2 Operations Procedure Group Supervisor - Responsible for maintaining this procedure current.

3.3 Operations Department Personnel - Responsible for performing operational activities in the manner described by the standards established by this procedure and associated procedures and OSMs.

4.0 PRECAUTIONS and LIMITATIONS 4.1 None 5.0 PREREQUISITES / INITIAL CONDITIONS 5.1 None INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 5 of 34 6.0 PROCEDURE 6.1 Operations Management and Leadership 6.1.1 Operations Management Team 6.1.1.1 The Director, Operations will establish an environment in which nuclear safety is an integral part of every operational decision. By consistent words and actions, he must be a credible role model to reinforce excellent safety culture behavior.

6.1.1.2 The Operations Department is in charge of the Plant. Specifically, the Shift Manager performs the role of Station Manager during assigned shifts and leads the Site by focusing Management attention on Operational and plant needs. As such, the Shift Manager possesses a deep and abiding concern for the success of all departments on Site, and is a role model for excellence in conduct, demeanor, and standards.

6.1.1.3 The Shift Manager has the ultimate command decision authority over all Plant activities and operations which could affect the safety of the general public, site personnel, and/or Plant equipment. To effectively implement this responsibility, he must maintain a broad perspective of Plant status and critical activities, not getting overly involved in any single activity. (Ref. 10.1.4.3, [CTS] Tech. Spec. 5.1.2 [ITS]

Delete) 6.1.1.4 The Shift Manager involves the operating teams in work schedule development prior to execution, and drives work schedule completion during execution week.

Involvement in work schedule development helps ensure the proper conservative philosophy is maintained, and that equipment is properly scheduled for repair commensurate with its importance to continued safe Plant operation.

6.1.1.5 The Shift Manager is an integral member of the Management team, and thus, is the conduit between Management and the on-shift operating team. The SM communicates and models Management expectations and standards, and intervenes when these expectations and standards are not being met. This intervention promotes consistency and continuous performance improvement further supporting public, personnel, and Plant safety.

6.1.1.6 Fostering teamwork, which includes free-flow of information, conservative approaches, and questioning attitudes, is critical for the Shift Manager to effectively implement the ultimate command decision authority.

6.1.1.7 The goal of Operations Management is to continually elevate the effectiveness of Operators, and promote error free performance. Operations Management will promote an attentive and thoughtful operating environment by defining expected behavior, emphasizing good operating practices, and providing the supervisory support needed to accomplish this goal.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 6 of 34 6.1.1.8 Operations Managers and Supervisors will periodically assess effectiveness in order to proactively manage by trying to anticipate problems before they occur. Identified problems are aggressively investigated to identify causes, direction is provided to implement actions to prevent recurrence, and department personnel are held accountable for achieving expected levels of performance.

6.1.1.9 Supervisory personnel will exhibit behavior and attitudes consistent with the high standards set by the Director, Operations. The work environment that results from the attitudes and behaviors of Operations personnel and the framework of policies and procedures fosters a safety culture for the entire Site organization.

6.1.1.10 The Director, Operations facilitates periodic meetings attended by Operations personnel, conducted in an open forum providing a supportive atmosphere for effective two-way (Line +-+ Management) communication, including:

Review of Operations Performance Standards and evaluation methods Division/Site/Company goals Management perspective Significant changes in the working environment Significant in-house or industry events Open discussion Change Management Operations Personnel development 6.1.2 Operator Fundamentals (Ref 10.2.5.4) 6.1.2.1 Operator fundamentals include performance skills and essential plant knowledge needed to operate effectively. Developing performance skills provides a quality of operation commensurate with our goal of ensuring nuclear, radiological, and personnel safety. A thorough understanding of plant design and principles of operation builds a solid foundation for improving operational performance and responding to adverse plant conditions.

6.1.2.2 Establishing standards in expected performance and knowledge promotes consistency in controlling the plant. The complexities of nuclear power require precise control, which demands ever-improving performance and deepening levels of knowledge. These are achieved through adherence to quality standards.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 7 of 34 6.1.3 Nuclear Safety Culture and Conservative Decision-Making (Ref. 10.1.4.2) 6.1.3.1 Safe nuclear power plant operation is based on the principle that each individual at SONGS accepts the unique and serious responsibility inherent in using nuclear technology. Every Operations employee has a challenging and crucial role in ensuring SONGS remains a safe and reliable source of electrical power.

6.1.3.2 An atmosphere that firmly establishes nuclear plant safety as an overriding priority in all aspects of plant operation is positively reinforced by all Managers and Supervisors. Nuclear plant safety issues receive the attention warranted by their significance, and other Site organizations such as Maintenance, Engineering and Work Control are aligned to support Operations in resolving these issues.

6.1.3.3 Management supports conservative decision-making and reinforces this philosophy through training, peer observation, and coaching. Plant and industry events are reviewed to gain lessons learned and to promote continuous improvement. The plant programs in place are designed to avoid activities that unnecessarily reduce nuclear safety margins.

6.1.3.4 Rigorous compliance to procedure use and adherence is the systematic process in decision-making which ensures we support a strong nuclear safety culture. WHEN conditions arise which are unexpected, or are outside the scope of normal operating conditions or procedures, THEN management promotes a culture which ensures Operations personnel do not proceed in the face of uncertainty. The Section for Problem-Solving/Decision-Making should be reviewed and Operators should place the Plant in a safe condition and obtain the appropriate guidance before proceeding.

(RCE 200694047, CA-15) 6.1.3.5 WHEN plant conditions are degrading, the Licensed Operator will identify action points as time permits (parameter, indicator(s), and value(s)) and ensure the following:

the reactor is manually tripped prior to a automatic reactor trip and when less than the available capacity of SBCS, the turbine is manually tripped prior to a automatic turbine trip.

6.1.3.6 On-shift licensed operators take the conservative actions necessary to maintain reactor safety. They are responsible for maintaining the core in a safe condition by deliberate control and by knowing the status of core reactivity, core cooling, and plant systems.

6.1.3.7 Actions that depart from Tech. Spec. or [CTS] LCS [ITS] TRM can be taken in an emergency when deemed necessary by the Shift Manager. Such actions shall only be taken after approval by the Shift Manager per 10CFR50.54(x) and 10CFR50.54(y). Refer to Attachment 4, Entry into 10CFR50.54(x) or 10CFR50.54(y) for actions. (NN 200345543) 6.1.3.8 Plant management supports on-shift personnel in making decisions and taking actions that ensure reactor and public safety.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 8 of 34 6.1.3.9 Economic and scheduling concerns are secondary to reactor safety. Senior plant management recognizes and assesses the impact of economic and scheduling decisions on reactor safety. A prudent margin to the plant's safety envelope is always maintained.

6.1.3.10 When Operators are faced with unexpected or uncertain conditions, then they must place the plant in a safe condition and must not hesitate to reduce power or trip the reactor.

6.1.3.11 WHEN time does not allow full understanding, risk should be minimized and the plant placed in a known safe condition. Operators must not feel a sense of haste. Haste can lead to an unintentional non conservative decision.

6.1.4 Operator Professionalism and Conduct Operations Managers and Supervisors foster an environment of continuous learning and improvement to achieve consistently high levels of performance. The following attributes of a professional Operator receive continuous attention:

PROTECTION OF THE PUBLIC - Operate in a manner that protects public safety above all other considerations. Recognize the responsibility to operate in a manner that merits public confidence.

CONSERVATIVE DECISION-MAKING - WHEN faced with unexpected or uncertain conditions, THEN Operators must place the plant in a safe condition, which may include a reduction of power or a trip of the Reactor.

REGULATORY COMPLIANCE - Adhere to all applicable regulations and procedures to ensure public protection and plant safety.

PERSONAL INTEGRITY - Demonstrate high principles and honesty in all aspects of one's job.

COMMITMENT TO EXCELLENCE - Strive to do the best job possible; the first time.

Begin with an end in mind. Plan the work; work the plan.

CONTROL OF PLANT OPERATIONS - Continually monitor plant conditions and aggressively resolve problems displayed by abnormal indications. Anticipate problems before they occur.

RESPONSIBILITY FOR ACTIONS - Assume accountability of one's actions and decisions. Assume ownership for the job and plant.

INDIVIDUAL KNOWLEDGE AND SKILLS - Continually work to improve job related skills and expertise.

PERSONAL INITIATIVE - Take action whenever conditions warrant. Recognize one's "response-ability" and use it to one's advantage.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 9 of 34 6.1.5 Operations Leadership Role in Site Activities 6.1.5.1 Operations personnel demonstrate ownership for all activities that affect Plant operation. The expectations established by Operations personnel have a profound effect on overall Site operation and the standards to which other Site organizations perform.

6.1.5.2 Operations Division will lead programs that are designed to maintain operational configuration control, to ensure Plant components and systems are always configured in a known manner and are capable of fulfilling their design functions.

6.1.5.3 Operations personnel will have a low tolerance for operator challenges in procedures and processes and are intolerant of degraded equipment conditions.

6.1.5.4 Operators lead the site to maintain the facility in excellent material condition to minimize equipment malfunctions that could challenge the operators during a transient.

6.1.5.5 All operations personnel must fully understand the integrated impact of equipment that is out of service and of proposed modifications or changes on safe plant operation. Concerns need to be communicated to senior Operations Management and appropriately resolved.

6.1.5.6 Operators will establish priorities for returning equipment to service based on a consideration of reduction in defense-in-depth and the need for operator compensatory actions.

6.1.5.7 Operational Focus is demonstrated when all SONGS personnel and organizations align on common goals and priorities that drive excellent material condition and support safe and reliable plant operations. Operations must take the lead in determining equipment priorities to ensure excellent material condition is achieved.

OSM-14, Operations Department Expectations, includes a list of equipment and plant systems with actions expected when plant conditions require timely resolution.

6.1.6 Personnel Safety Culture (SONGS Industrial Safety Pocket Guide) 6.1.6.1 Individual safety must always be the number one priority during task performance.

6.1.6.2 The key elements of a questioning attitude also apply to worker safety.

Understanding safety rules and enforcing safe work practices throughout the Plant is a basic expectation and applies to all Operations employees. IF a concern for the safety of the workers is involved, THEN the job should be stopped until the concern is resolved satisfactorily.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 10 of 34 6.2 Control Room Operations 6.2.1 Control Room Access 6.2.1.1 Control room access is limited to persons on official business only. The number of personnel in the Control Room is limited to reduce distractions to the operators.

Control Room supervisors rigidly uphold management expectations for controlling access to the main Control Room. (DCE 200403800) 6.2.2 Control Room Board Monitoring 6.2.2.1 The Control Room is the most important plant operating station and is the central coordination point for station operations. It is important to manage Control Room activities so operators are not distracted from properly monitoring plant parameters.

6.2.2.2 Operators monitor control board indications and alarms frequently and act promptly to determine the causes of and corrections for abnormalities. Emphasis is placed on closely monitoring, investigating, and trending to detect problem situations early.

6.2.2.3 Computer-generated graphics can be used effectively to provide maximum trending and monitoring capability to the Control Room operators. However, the use of computer screen monitoring capabilities will be balanced with monitoring of installed main control board meters and recorders.

6.2.2.4 The number of concurrent evolutions affecting control board indications and alarms will be limited to prevent compromising the operators' abilities to detect and respond to abnormal conditions.

6.2.3 Operation / Manipulation of Plant Equipment 6.2.3.1 The manipulation of systems and components in the plant is the responsibility of the Operations Division. Delegation of these responsibilities to other divisions' personnel is allowed, provided adequate controls are in place to ensure proper operation and maintenance of plant status control.

6.2.3.2 The Shift Manager may suspend the authority of other divisions' personnel to operate plant components due to misoperations.

6.2.3.3 Trainees who operate equipment are supervised and controlled by the qualified operators who normally perform the operations.

6.2.3.4 Appropriate error reduction techniques will be used while manipulating plant controls and components. Error-prevention techniques used include, but are not limited to, the following: self-checking (STAR), placekeeping, peer-checking, three-way communications, use of error reduction tools.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 11 of 34 6.2.4 Administrative Duties of Control Room Personnel 6.2.4.1 Shift supervisory personnel, including the Shift Manager will remain in a supervisory oversight role by directing plant operations, the control room crew, and especially the assistant control operators and should minimize administrative duties while filling this role.

6.2.4.2 Administrative duties assigned to control room licensed operators (SM, SRO, RO) should not interfere with their abilities to monitor plant parameters or provide supervisory oversight of the crew.

6.2.4.3 Any administrative responsibilities, assigned to the SM are reviewed and approved by the Director, Operations. This review shall not be delegated. (Ref. 10.1.4.3) 6.2.4.4 If one control room operator is distracted by an administrative task, other control room operators assume responsibility for monitoring the unit.

6.2.5 Reactivity Management 6.2.5.1 The SONGS Reactivity Management Program is established, defined and implemented in S0123-RX-1 and S0123-XV-91. OSM-14, Operations Department Expectations provides the implementation practices specific to the Operations Department.

6.2.5.2 Reactivity manipulations are only performed when directed by approved plant procedures, are only performed with Control Room Supervisor or Shift Manager permission, are accomplished with supervisory oversight of the evolution and the reactor operator is free from distractions and other duties.

6.2.5.3 Reactivity manipulations are made in a deliberate, carefully controlled manner while the reactor is monitored to ensure the desired response is attained.

6.2.5.4 Reactivity manipulations are made while redundant instrumentation is observed to determine the effects of the reactivity manipulation.

6.2.5.5 Manipulation of either CEDMCS (CEA Movement) or CVCS for the purpose of Boration or dilution must be under the direct control of a Licensed Operator. The single exception is an Operator-In-License training who is under the direct supervision of a Licensed Operator. (1 OCFR 50.54) 6.2.5.6 Operation of equipment that may indirectly (other than CEDMCS or CVCS) affect the power level or reactivity of the reactor shall only be accomplished with the knowledge and consent of a Licensed Operator. (10CFR 50.54)

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 12 of 34 6.2.6 Control Room Communications, Briefs and Updates 6.2.6.1 The Shift Manager is responsible for ensuring communications, including the use of three-way communications, are in accordance with established standards.

6.2.6.2 Licensed Operators are responsible for making appropriate notifications when changing Plant status, including:

Prior to activities that could change radiation levels Prior to starting/stopping Plant equipment and energizing/de-energizing electrical switchgear (personnel safety concern)

For mode changes, criticality, etc.

To alert Plant personnel of an unexpected event or condition, (local area evacuations, Fire Department response, Emergency Plan assembly, etc.)

6.2.6.3 It is acceptable and encouraged for control room operators to have discussions regarding plant status or evolutions. It is important that the discussion phase have a clear ending and the communication phase have a clear beginning where three-way communications is required to verify understanding.

6.2.6.4 There are two methods for broadcasting information to the Crew, a brief (to ensure all team members get the same message) or a update (a short message of importance).

6.2.6.5 Direction is never issued during discussions and Updates. IF during a Brief, assignment of roles and responsibilities need to be changed or made, THEN 3-way communication is required to ensure understanding.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 13 of 34 6.2.7 Control Room Command Function 6.2.7.1 A Licensed Operator shall be responsible for the Control Room Command Function for both Units.

6.2.7.2 The individual holding the Control Room Command Function shall have the following responsibilities:

Maintain COMMAND AND CONTROL by establishing intent - Defining and prioritizing tasks, operations or activities, which includes the desired objective and the risk involved; determining roles, responsibilities, and relationships - To enable, encourage, and constrain specific types of behavior to perform the actions needed to accomplish the intended task, operation or activity; establishing rules and constraints - Task, operations or activities will be conducted through rigorous procedure use and adherence, clearly understanding each action and expected results, and stopping when an unexpected condition or response is obtained; monitoring and assessing the situation and progress - Being aware of changes in plant conditions and the appropriateness and timeliness of the response to those changing conditions.

COMMAND AND CONTROL is defined in Section 9.0 and discussed in detail in OSM-12, Operator Fundamentals. (CAPR 201045203 CA0001)

Build TEAMWORK by having operational crews work together to resolve problems, regardless of ownership, through establishing clear goals which are specific, measurable, achievable, and timely; clarifying roles and responsibilities and by clarifying the process of decision-making, and requesting cross-organizational collaboration. TEAMWORK is defined in section 9.0 and discussed in detail in OSM-12, Operator Fundamentals. (CAPR 201045203 CA0001)

Maintaining Control Room decorum by providing oversight, enforcing expectations, managing resources, and monitoring Crew performance.

END OF SECTION INFORMATION USE

Sa123-0-Al S0123-0-Al Rev 41 Page 14 of 34 6.2.8 Shift Manning 6.2.8.1 Shift manning will be maintained in accordance with the Unit Technical Specifications Sections 5.2.2, Unit Staff, and 5.3, Unit Staff Qualifications.

6.2.8.2 Shift manning is detailed in Attachment 2.

6.2.9 Annunciator Response 6.2.9.1 The operating crew will be aware of the Control Room annunciator status, including alarms that are totally disabled, alarms with individual inputs disabled, computer-generated alarms that are taken out of scan, those with temporarily changed setpoints and multiple input alarms that do not reflash when more than one input is activated.

6.2.9.2 WHEN alarms are impaired, Operators will establish alternate means of indication to verify compliance with technical specifications and safety function and monitor equipment parameters for abnormal conditions that would be masked by deficient or altered alarms.

6.2.9.3 The operating crew will be aware of and anticipates annunciators that alarm due to testing or as a result of normal operating conditions or work activities.

6.2.9.4 The operating crew will respond to annunciators per applicable alarm response procedures.

6.2.9.5 During normal operations, when an alarm annunciates, or an alarm condition clears, all conversations in the Control Room should stop until the alarm has been acknowledged or reset.

6.2.9.6 If an in-solid alarm condition is due to a parameter outside its normal operating band and the potential for equipment degradation or reduced operating margin exists that could reasonably lead to any of the following:

Significant Personnel Safety Hazard Reactor Trip Loss or trip of equipment important to nuclear safety.

Inadvertent entry into a Tech Spec LCO of 7 days or less Significant loss of generation then the SRO Ops Supv is responsible to evaluate the need for initiation of OSM-1 14, Adverse Condition Monitoring Plan.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 15 of 34 6.3 Operations Field Activities 6.3.1 Operators assigned the task of rounds are responsible to verify that operating and standby equipment operates within normal parameters. The operators should note degrading equipment and environmental conditions such as water and oil leaks, burned-out light bulbs, and changes in building temperature or air circulation. Operators will inform control room personnel of problems noted and initiate an NN. Refer to OSM-14, Operations Expectations, Section for Operations Field Activities and OSM-5, Operator Rounds for detailed component inspection criteria.

6.3.2 Operators will identify actual and potential equipment problems, and conditions that could impact the function of the equipment, at a point early enough for supporting organizations to correct the problems.

6.3.3 The operator who conducts field inspections is the "eyes and ears" for the control room team. As such, the organization shall treat the concerns raised by the field operator seriously.

6.3.4 The field operator is regarded as the owner of the assigned equipment and area until relieved of that responsibility; this means the operators assigned are aware of scheduled work activities in their area and have an input regarding the level of priority placed on equipment and condition deficiencies identified.

6.3.5 Field operators will identify and correct other adverse conditions related to personnel and equipment protection, such as personnel safety hazards (e.g., open electrical panels, unlocked high radiation doors, and missing labels), placement of material and equipment that would impede abnormal or emergency response, accumulation of combustibles and trash, improper or inadequate storage of chemicals and equipment, and clogged floor drains.

6.3.6 Operators will maintain radiation dose as low as reasonably achievable (ALARA) and ensure the number of contaminated areas has a minimal impact on the routine inspections conducted by field operators.

6.3.7 Unauthorized breach of a Hazard Barrier may compromise the safety of the plant, plant personnel and the public. Therefore, all barriers at SONGS 2 & 3 are assumed to be Hazard Barriers until evaluation by Engineering concludes otherwise. Refer to S023-XV-4.500, Control of SONGS 2 and 3 Barriers, for more information.

INFORMATION USE

S0123-0-Al Sa123-0-Al Rev 41 Page 16 of 34 6.4 Shift Activities 6.4.1 Operator Log Keeping 6.4.1.1 Operations department logs, Unit 2, Common, and Unit 3, contain an official chronological listing of events relevant to unit operation occurring on a given shift, such as Shift assignments, Personal safety events, Mode Changes, Power Changes, Reactivity Changes, LCO Entry and Exit, Entry into Emergency or Abnormal procedures, Implementation of the Emergency Plan, Major equipment status changes or events deemed necessary by Control Room personnel to aid in shift turnover, Offsite notifications, Significant operational decisions, Performance and results of a risk assessment to address emergent condition(s).

6.4.1.2 Logs should be legible and entries should be made in a narrative manner so that they may be easily read and understood.

6.4.1.3 Information should be captured in the narrative log in sufficient detail to accurately categorize system and component operability and availability.

6.4.1.4 To aid in event reconstruction, as much significant information as possible should be logged during emergencies and abnormal or unexpected events. A rough log may be kept provided that entries are made in the narrative log as soon as possible.

6.4.1.5 Logkeeping does not take precedence over control and monitoring of the plant.

6.4.1.6 Shift supervision will review the logs to ensure that the logs are accurate and appropriate.

6.4.1.7 Operations management will periodically review the logs to determine if log keeping is in accordance with operations department expectations.

6.4.1.8 All log entries relating to operability shall have SRO concurrence in the log with name of SRO providing Independent Verification.

6.4.1.9 The requirements in OSM-14, Section for Operator Log Keeping should be used when making log entries.

6.4.1.9.1 The following attributes should be answered when making log entries:

Who What When Where Why How 6.4.1.9.2 OSM-14, Attachment for Log Entry Events / Activities provides an example of what each term above means and minimum acceptable criteria.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 17 of 34 6.4.2 Operator Turnovers and Shift Relief Status Sheets 6.4.2.1 Shift relief will be conducted in a professional, disciplined and formal manner.

All turnovers should follow a similar protocol. The turnovers transfer specific legal responsibilities to qualified individuals.

6.4.2.2 No Operator shall assume a watchstation unless they are physically and mentally fit to competently perform the responsibilities, nor shall any person permit their relief to assume the shift if any doubt exists concerning the reliefs alertness, coherence, and capability of performing their assigned duties.

6.4.2.3 Any Operator who is absent from work due to illness for 3 or more consecutive scheduled shifts shall obtain a release from his/her personal physician prior to returning to work. IF the Operator is licensed, and the absence is for > 5 consecutive scheduled shifts, THEN a release from the Edison Medical Department evaluating the Operator's abilities to perform his/her duties is required.

6.4.2.4 The Shift Manager may request a release following any illness if there is a question regarding fitness for duty.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 18 of 34 6.4.3 Work Process 6.4.3.1 The Shift Manager or licensed designee authorizes all maintenance and testing activities on equipment important to safety that affects plant operations or that changes control room indications or alarms.

6.4.3.2 On-line maintenance is planned and scheduled for preventive and corrective maintenance with minimum overall safety impact and minimum out-of-service times for important systems or equipment. As a result, system and overall plant reliability are increased, equipment and system deficiencies that could affect operations are reduced, and outage time and scope are reduced.

6.4.3.3 Deviating from the schedule is error likely. Work that is unscheduled requires Supervisor or SM concurrence, and maintenance is held accountable for working in accordance with the pre-approved schedule. Operations personnel are held accountable for supporting maintenance as scheduled. The SM must be notified prior to turning away any scheduled work task.

6.4.3.4 The Supervisor-Plant Operations or SM is responsible to control plant configuration.

6.4.3.5 All work activities, including emergent work, are factored into the risk assessment for impact and any reduction in the margin to safety. High risk activities are given a higher level of review that may result in the work being deferred.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 19 of 34 6.5 Human Performance and Error Reduction 6.5.1 Monitoring and Assessing Performance 6.5.1.1 Operations management is responsible to develop goals and objectives for the Operations department that support the station and corporate goals and address any performance improvements needed. The goals are realistic, measurable, achievable and challenging with the purpose to improve performance.

6.5.1.2 Operations Managers and Supervisors participate in regular performance observations of Operations activities through a variety of methods, such as plant tours, training observations, participation in shift activities, and by participating in the Leadership Observation Program (LOP). These observations are used as a vehicle for coaching and communicating high standards of personal performance to each individual. By continuously monitoring personnel performance and measuring it against goals and standards, causes for performance shortfalls can be identified and corrected.

6.5.2 Evaluations of Crew Performance 6.5.2.1 The Shift Manager is responsible to communicate the goals and objectives to the crew members ensuring an understanding of the tie to station and corporate goals.

6.5.2.2 The Shift Manager is responsible for reviewing all trend data and developing corrective actions for performance shortfalls. Crew performance will be reviewed during crew MRM's with Operations management.

6.5.3 HU Events / Error Investigation 6.5.3.1 The investigation of HU issues can identify detrimental conditions that, if left uncorrected, could affect plant operations. The rigor, scope, and depth of investigations reflect the potential safety significance and consequences, as well as the likelihood that the event will recur. Establishing interim actions in a timely manner may be necessary to prevent recurrence of a similar event.

6.5.3.2 The Shift Manager will ensure information is collected promptly after an event.

Being timely ensures the availability of personnel who observed or participated in the event and their insights into the plant response and actions leading up to the event are not forgotten.

6.5.3.3 WHEN an Operations HU event or error is determined or suspected, THEN initiate OSM-8, Error Disposition Instruction, and refer to S0123-XV-HU-1, Human Performance Program.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 20 of 34 6.5.4 Problem-Solvinq/Decision-Makinq/Risk Management 6.5.4.1 Problem-solving and decision-making will always have a bias towards safety.

See Section 6.1.3, Nuclear Safety Culture and Conservative Decision-Making, for guidance.

6.5.4.2 All activities are evaluated for risk. Risk can be evaluated quantitatively using the Safety Monitor per S023-XX-10, Maintenance Rule Risk Management Program Implementation. All activities are evaluated qualitatively, per S023-XX-8, Integrated Risk Management. The qualitative assessment will determine task risk level, identify the tools to manage the risk and perform an integrated risk assessment. Refer to OSM-14, Operations Department Expectations, Section for Risk Assessment, for those activities requiring Operations Management and Senior Leader observation.

6.5.4.3 Problem-solving must follow a formal consistent process to ensure a consistent outcome. Problem-solving many times includes troubleshooting. General troubleshooting guidelines are addressed in S023-XV-2, Troubleshooting Plant Equipment and Systems.

6.5.4.4 Decision-making is a part of normal daily activities. For decisions beyond normal daily activities refer to S0123-XX-19, Operational Decision-Making Process.

(NN 200927470)

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 21 of 34 6.6 Human Performance Tools 6.6.1 Use of Human Performance Tools is absolutely essential to the success of SONGS.

Use of these tools can greatly reduce and/or eliminate Human performance errors. It is the responsibility of all Operations personnel to apply these tools appropriately during the performance of their work. (AR 040801655-02) 6.6.2 S0123-XV-HU-2, Human Performance Tools applies to all SONGS workers. Refer to OSM-14, Operations Department Expectations for Operations specific applications of HU tools.

6.7 Operator Knowledge and Development 6.7.1 Management Standards for Traininq 6.7.1.1 Training and qualification programs will be systematically used to address the needs of operations personnel and provide the foundation for their continuous development and improvement.

6.7.1.2 Training and qualification programs for operations personnel will develop, maintain and enhance the knowledge and skills needed by nonlicensed operators, reactor operators, senior reactor operators, shift technical advisors, and Shift Managers to effectively perform operations activities in the plant. Training develops the abilities of operations personnel to respond appropriately to unusual incidents and to anticipate and prevent such events.

6.7.2 Operator Training Program 6.7.2.1 Operations department managers and Shift Managers are responsible for the training and qualification of operations personnel. This requires close coordination with the training department to:

establish and maintain course content and emphasis, to determine and support training schedules, to accomplish on-the-job training and evaluation, to observe training activities, and to provide feedback to adjust course content and emphasis.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 22 of 34 6.8 Procedure Use and Adherence (SO123-XV-HU-3, S0123-0-A3) 6.8.1 The plant shall be operated in accordance with applicable station procedures.

Operations procedures provide appropriate direction such that the plant is operated within its design bases and support safe operation of the plant.

6.8.2 S0123-XV-HU-3, Written Instruction Use and Adherence requires Continuous Use procedures to be in the performer's presence or within view. This requirement is referred to as "in-hand" by Operations. Documents required to be "in-hand" do not need to be physically in the performers' hand, but need to be close enough to support the performance of STAR (matching label to document) and circle/slash.

6.8.3 Activities designated as Skill-of-the-Craft do not require procedure in hand for component manipulation. OSM-14, Operations Department Expectations includes a list of routine tasks that have been designated Skill-of-the-Craft.

6.8.4 Operators are expected to take time to prepare for an activity by reviewing the procedure and, regardless of the use category assigned (continuous, reference, information), operators must understand completely the procedure steps to ensure correct performance of an activity or an evolution.

6.8.5 The Manager, Plant Operations (or designee) approves all operator aids to confirm they are necessary and correct. Operator aids may supplement, but are not used in lieu of, approved procedures, station tagging, or plant labeling.

6.8.6 During routine plant inspections, operations personnel will review operator aids to confirm they are accurate and approved.

6.8.7 Procedures are reviewed prior to use. This review is expected to identify any needed modifications and the modifications will be made prior to use of the procedure.

6.8.8 Perform procedure steps as written, in the sequence specified, except when the written instruction or approved processes specifically allows deviation.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 23 of 34 6.9 Administrative 6.9.1 Active License Status Requirements NOTE

1) Augmenting shift crews with Licensed Operators in order to maintain proficiency in the performance of Licensed Operator functions is permitted. However, more than one additional watchstander at each position is not acceptable with respect to the proficiency issue. (Ref. NUREG - 1262, page viii)

2) An Operator who becomes active or maintains active license requirements in one quarter will remain active for the following full quarter.

3) Operations Forms may be found under SAP Express > Document Search >

Controlled Forms >Document number *OP(123)*.

4) Licensed Operators (SRO, RO) are responsible for maintaining the necessary requirements to keep an active license.

6.9.1.1 An RO or SRO must be actively performing the functions of an RO or SRO to maintain their license active. To maintain active status, the licensee shall actively perform the functions of a Certified Operator, Supervisor, or SM a minimum of seven 8-hour or five 12-hour shifts per calendar quarter. Ref. 10CFR55.53 (e) and NUREG-1262 6.9.1.2 The minimum number of shifts that must be worked to keep a license active shall be seven 8-hour or five 12-hour shifts per calendar quarter. Tracking is performed per, or other formal process. (Ref. 1 OCFR55.53e) 6.9.1.3 For bargaining unit employees with an SRO license, in order to maintain an SRO license active, the operator must stand a minimum of one complete watch (8-hour or 12-hour shift) per calendar quarter in a credited SRO-only position. The remainder of complete watches required in a calendar quarter may be performed in either a credited SRO or RO position. (NN 200006689) 6.9.1.4 IF a licensed SRO stands all of their required proficiency watches in an SRO-only supervisory position, THEN the RO portion of the license is still considered active.

(NN 200006689) 6.9.1.5 Inactive licenses may be returned to active status as delineated in OP(123) 25, Reactivation of Inactive NRC License. The licensee must be current in the requalification program before returning to duty and complete a minimum of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of shift functions under the direction of an RO or SRO, in the position to which the individual will be assigned (i.e., RO working as Certified Operator and SRO working as a Supervisor-Plant Operations or SM). The 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> must include a review of the shift turnover sheets and a complete tour of the plant under the direction of an individual actively licensed at a level at least equal to the license being reactivated.

Badging records, watches, and plant tours are needed to support reactivation. Any hours working as a licensed trainee also count towards the 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. For example, an SRO coming back on shift and standing SM-T watches count. (Ref. 10CRF55.53f, NRC 71111.11, Inspection of the License Operator Requalification Program)

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 24 of 34 6.9.1.6 Reactivation of an SRO license, for refueling activities only, shall be completed as delineated in OP(l 23) 26, Limited Reactivation of Inactive NRC License for Refueling Activities. Badging records, watches, and/or plant tours are needed to support reactivation per NRC 71111.11, Inspection of the License Operator Requalification Program. Attach Log entries for reactivation watches or plant tours and security records supporting in-plant time to the OP(1 23) 26 form.

6.9.1.7 Inactive licensed individuals shall not stand licensed watches as specified in the Technical Specifications and Licensee Controlled Specifications.

6.9.1.8 CC-3, Qualification Tracking, is used to track watchstation qualification. Licensed Operators are responsible for knowing his qualifications and only standing watches he is qualified to stand.

6.9.1.9 Inactive licensed individuals may perform all non-Licensed duties which they are qualified to perform and stand any non-licensed watchstation for which they are qualified and proficient.

7.0 ACCEPTANCE / FUNCTIONAL CRITERIA 7.1 None 8.0 RETENTION / RECORDS 8.1 Record Retention guidelines are described in S0123-0-A3, Procedure Use.

8.2 Completed Operator Logs are to be maintained at/near the user location for three months, and THEN transmitted to CDM-SONGS by the Office Assistant.

8.3 Operations Division Files should be transmitted to CDM-SONGS, as directed by the Work Control Supervisor on a routine basis.

8.4 The Office Assistant should retain copies of completed OP(123) 27, Active License Tracking, to man the shifts for the upcoming quarter.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 25 of 34 9.0 DEFINITIONS 9.1 CONSERVATIVE ASSUMPTION - A hypothesis, theory, supposition, or premise that is accepted as true and can be verified through supporting documentation; design criteria accepted as true or conservative in order to bound inputs-(Alternatively, an unverified assumption is an assumption that has not or cannot be validated or trusted as correct without additional data or testing.)

9.2 COMMAND AND CONTROL - The exercise of authority and direction by the SRO with the Control Room command function over the on-shift crew in the accomplishment of maintaining nuclear safety. Command and control functions are performed through an arrangement of personnel, equipment, communications, facilities, and procedures employed in planning, directing, coordinating, and controlling operations in the accomplishment of maintaining nuclear safety.

9.3 CONTINGENCY PLAN - Plan devised for a specific situation, e.g., performing CRITICAL STEP(s),

performing a reactivity affecting activity or when the risk of things going wrong is not acceptable.

Contingency planning is unique to each activity, providing preventive measures, recovery strategies, and technical considerations appropriate to the activity.

9.4 CRITICAL STEP - A written instruction step, series of steps, or action that, if performed improperly, will cause irreversible harm to plant equipment, to people, or will significantly impact plant operation. [INPO 09-004]

9.5 DIRECT OVERSIGHT - The sole and primary function of a supervisor or manager while supervising an activity. That is, the supervisor or manager will respond to request only from those performing the activity and not be distracted by answering the telephone or being involved or conducting concurrent activities.

9.6 OFF-NORMAL CONDITION - Any one of a group of processes or system variables, e.g., flow, pressure, temperature, voltage, current, which cause some or all of the remaining conditions in the process or system to become abnormal. For example, an alarming annunciator indicates a process or system variable setpoint has been exceeded.

9.7 OVERSIGHT FUNCTION - Oversight is the critical function provided by supervision and management to the proper operation of the plant in accordance with approved programs, processes, and procedures. It is the job of oversight to assure that STANDARDS and EXPECTATIONS are adhered to and that supervision and management re-enforces and measures the effectiveness of the individuals in fulfilling those STANDARDS and EXPECTATIONS.

9.8 PLACE KEEPING - Physically marking steps in a written instruction to prevent the omission or duplication of the steps to maintain an accounting of steps in progress, steps completed, steps not applicable, and steps not yet performed. [INPO 09-004]

9.9 REACTIVITY AFFECTING ACTIVITY - Is a task or plant manipulation, which changes or has the potential to change reactor power or power indication, RCS temperature, pressure, boron concentration, or CEA position(s).

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 26 of 34 9.10 STANDARDS - Written description of programs, processes, and procedures established by Management as a rule to measure quality or performance.

For example:

The standard for configuration and control is the plant will always be aligned in accordance with approved procedures.

The standard for reactivity management is all activities ensure core reactivity and stored nuclear fuel (where the potential for criticality can occur) are monitored and controlled consistent with fuel design and operating limits.

" The standard for plant control is all activities ensure strict adherence to approved procedures as intended and as written, results of actions are clearly understood, and back-out criteria are established.

9.11 EXPECTATIONS - Observable behaviors and attitudes which exhibit responsibility and accountability to ensure all activities performed satisfy STANDARDS.

For example:

A. In part, the expected behavior for an operator when changing plant configuration is:

Completing a Pre-Job Brief Always follow, as written and intended, an approved operating instruction, Use the HU Tools for self-checking: STAR, flagging, Positive Component Verification, Use Place Keeping STOP with an unexpected condition or response and promptly notify the CRS.

B. In part, the expected behavior when performing a reactivity affecting activity is:

Completing a reactivity brief with the Shift Manager in attendance.

The activity is conducted with a procedure in hand following the requirements for procedure use and adherence Manipulations are clearly identified, fully understood and closely monitored to verify the expected magnitude, direction, and effects are realized STOP with an unexpected condition or response and promptly notify the CRS.

C. In part, the expected behavior when performing a plant maneuver is:

Completing a Pre-Job Brief Always follow, as written and intended, an approved operating instruction, Each team member understands parameters monitored and expectations for the instrument response.

STOP with an unexpected condition or response and promptly notify the CRS.

9.12 PROMPT AND PRUDENT - An action URGENT and NECESSARY to address an adverse trend in order to place the plant in a stable condition; and the action is acceptable as correct without first reference to a supporting procedure. [CAPR 201045203-CA0001]

9.13 TEAMWORK - A sharing in the responsibility and accountability in planning, directing, coordinating, and controlling operations in the accomplishment of maintaining nuclear safety.

INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 27 of 34

10.0 REFERENCES

/ COMMITMENTS 10.1 Implementing Reference 10.1.1 Commitments 10.1.1.1 None 10.1.2 Corrective Actions to Prevent Recurrence (CAPR) 10.1.2.1 None 10.1.3 Procedures 10.1.3.1 SO123-0-A3, Procedure Use 10.1.3.2 SO123-0-A7, Notification and Reporting of Significant Events 10.1.3.3 S0123-VIII-10, Emergency Coordinator Duties 10.1.3.4 SO123-XV-HU-1, Human Performance Program 10.1.3.5 SO123-XV-HU-3, Written Instruction Use and Adherence 10.1.3.6 S0123-XX-19, Operational Decision Making Process 10.1.3.7 SO23-XV-2, Troubleshooting Plant Equipment and Systems 10.1.3.8 SO23-XV-4.500, Control of SONGS 2 and 3 Barriers 10.1.3.9 S023-XX-8, Integrated Risk Management 10.1.3.10 S023-XX-10, Maintenance Rule Risk Management Program Implementation 10.1.4 General 10.1.4.1 CC-3, Qualification Tracking 10.1.4.2 D-003 ISS3, Nuclear Safety Culture and the Nuclear Organization's Priorities 10.1.4.3 D-004, Management Responsibilities for the Units 2/3 Control Room Command Function 10.1.4.4 OP(123) 27, Active License Tracking 10.1.4.5 OSM-1, Operations Dictionary 10.1.4.6 OSM-8, Error Disposition Instruction 10.1.4.7 OSM-12, Operator Fundamentals 10.1.4.8 OSM-14, Operations Department Expectations INFORMATION USE

S0123-0-Al SO123-0-Al Rev 41 Page 28 of 34 10.2 Developmental References 10.2.1 Commitments 10.2.1.1 Units 2 and 3 Technical Specifications 10.2.1.2

[CTS] Licensee Controlled Specifications (LCS)

[ITS] Technical Requirements Manual (TRM) 10.2.1.3 Units 2 & 3 UFSAR, Section 13 10.2.1.4 Unit 2 and 3 Updated Fire hazards Analysis, San Onofre Nuclear Generating Station 10.2.1.5 10CFR50, Domestic Licensing of Production and Utilization Facilities 10.2.1.6 10CFR50 Appendix R, Section III.G 10.2.1.7 10CFR50 Appendix R Section III.H 10.2.1.8 10CFR55, Operators' Licenses 10.2.1.9 Reg. Guide 1.114, Guidance on Being Operator at the Controls of a Nuclear Power Plant 10.2.1.10 NUREG 0737, Post TMI Requirements, I.C.4 10.2.1.11 Reg. Guide 1.78, Control Room Habitability Requirements 10.2.1.12 10CFR26, Subpart I, Managing Fatigue 10.2.1.13 SONGS Emergency Plan, Section 5.0 10.2.2 Corrective Actions to Prevent Recurrence (CAPR) 10.2.2.1 CAPR 201045203 CA0001, Definition of Prompt and Prudent and consolidation of Command and Control and Teamwork Standards 10.2.3 Procedures 10.2.3.1 SO123-0-A2, Operations Division Personnel Responsibilities 10.2.3.2 SO1 23-0-A6, Routine Equipment Operations 10.2.3.3 SO123-IT-1, Infrequently Performed Tests or Evolutions Control Program 10.2.3.4 S0123-IV-4.4, Lock and Key Issue, Control and Accountability 10.2.3.5 S0123-IV-6.8.5, Security Compensatory Measures 10.2.3.6 S0123-RX-1, Reactivity Management Program 10.2.3.7 SO123-TN-1, Nuclear Organization Training 10.2.3.8 S0123-XV-6.1, SONGS Fatigue Management 10.2.3.9 S0123-XV-6.2, Work Hour Controls 10.2.3.10 S01 23-XV-6.3, Work Hour Waivers 10.2.3.11 S0123-XV-27, On-The-Job Training and Task Performance Evaluation Program 10.2.3.12 S0123-XV-73, SONGS Sign Production Process 10.2.3.13 SO123-XV-91, Reactivity Management Implementation 10.2.3.14 SO123-XV-HU-2, Human Performance Tools 10.2.3.15 SO123-XXI-1.11.4, Non-Licensed Operator Training Program Description 10.2.3.16 S023-5-1.7, Power Operations 10.2.3.17 S023-6-29, Operation of Annunciators and Indicators 10.2.3.18 S023-13-28, Rapid Power Reduction (RPR) 10.2.3.19 S023-XV-70, Operator Licenses 10.2.3.20 S023-XXI-1.11.5, Reactor Operator/Senior Reactor Operator Initial License Training Program Description 10.2.3.21 S023-XX1-1.11.12, Shift Technical Advisor Training Program Description 10.2.3.22 S023-XXI-1.11.20, Senior Control Room Supervisor (SCRS)/Shift Manager (SM)

Training Program Description 10.2.3.23 S023-XXX-VII 13, Control of the Plant Physics Data Books and Reactor Engineering Data Transmittals INFORMATION USE

S0123-0-Al S0123-0-Al Rev 41 Page 29 of 34 10.2.4 Web Links & Forms 10.2.4.1 AD(123) 10, Termination/Change of Status Notification 10.2.4.2 AUD-1, Operations Records Recommended for Internal Audit 10.2.4.3 NOA-1, Non-Security Key Index 10.2.4.4 NOA-2, Non-Security Key Issue Log 10.2.4.5 OP(123) 1, Status Control Form 10.2.4.6 OP(123) 24, Extra Person on Shift (EPOS) Documentation 10.2.4.7 OP(123) 25, Reactivation of Inactive NRC License 10.2.4.8 OP(123) 26, Limited Reactivation of Inactive NRC License for Refueling Activities 10.2.4.9 OP(123) 28, Procedure Modification Permit 10.2.4.10 OP(123) 29, Abnormal Evolution 10.2.4.11 OP(1 23) 34, Annunciator Compensatory Actions 10.2.4.12 OP(123) 35, Annunciator Setpoint Change 10.2.4.13 OSM-6, Operations Department Human Performance Tools 10.2.4.14 OSM-9, Standard EOI and Off-Normal Event Good Practices and Strategies 10.2.4.15 PRO-10, Operator Aid Action Request 10.2.4.16 SE(123) 41-2, Key Issue Log 10.2.5 General 10.2.5.1 INPO 01-002, Guidelines for the Conduct of Operations at Nuclear Power Stations 10.2.5.2 INPO 06-006, Guidelines for Effective Reactivity Management 10.2.5.3 INPO 09-004, Procedure Use & Adherence 10.2.5.4 INPO IER 11-3, Weaknesses in Operator Fundamentals 10.2.5.5 NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management - SONGS Assessment; dated December 1992. (101-458) 10.2.5.6 INPO, Principles for a Strong Nuclear Safety Culture, November 2004 10.2.5.7 Priority 2 Reading 2-06-038, dated 4/24/06;

SUBJECT:

Setting Expectations, author Dennis Wilcockson 10.2.5.8 NRC Information Notice 2006-09, "Performance of NRC Licensed Individuals while On Duty with Respect to Control Room Attentiveness" 10.2.5.9 NEI 06-11, Managing Fatigue Rules at Nuclear Power Reactor Sites 10.2.5.10 SER 3-05, Weakness in Operator Fundamentals 10.2.5.11 NUREG-1 262, Answers to Questions at Public Meetings Regarding Implementation of Title 10, Code of Federal Regulations, Part 55 on Operators' Licenses, page viii 10.2.5.12 WANO SOER 2007-01, Reactivity Management 10.2.5.13 RIS07-29, Clarified Guidance for Licensed Operator Watch-Standing Proficiency INFORMATION USE

S0123-0-Al ATTACHMENT 1 Rev 41 S0123-0-Al Page 30 of 34 THIS ATTACHMENT IS NOT USED INFORMATION USE ATTACHMENT 1 Page 1 of 1

SO 123-0-Al ATTACHMENT 2 Rev 41 SO123-0-Al Page 31 of 34 MINIMUM OPERATIONS SHIFT CREW COMPOSITION 1.0 ERO Minimum Crew Composition is addressed in SONGS Emergency Plan 2.0 Non-ERO Total Minimum Crew Composition when BOTH Units are Defueled: [1] [3]

Position Number Required Reference SRO / CFH 1

10 CFR 50.54 (m)(2)(i)

RO 2 [2]

10 CFR 50.54 (m)(2)(i)

Non-Licensed 3 [2]

Tech Spec 5.2.2.a Footnotes

[1]

Except for the Shift Manager, the shift crew composition may be one less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift crew members provided IMMEDIATE ACTION is taken to restore the shift crew composition to within the minimum requirements. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewperson being late or absent. (Tech. Spec. 5.2.2.c, [CTS] LCS 5.0.100.1.1 [ITS] Delete)

[2]

RO Licensed individuals may fill both RO and Non-Licensed positions.

[3]

A Designated Shift Communicator is required at all times to support the Emergency Plan function of Shift Communicator.

INFORMATION USE ATTACHMENT 2 Page 1 of 1

SO123-0-Al ATTACHMENT 3 Rev 41 S0123-0-Al Page 32 of 34 DOCUMENTED ON-SHIFT HOURS (1 OCFR55.57)

Operator Name Date Period of this report: from to Date Date During the period of this report the above named individual was on shift and thus operated the facility approximately Hours The method used to approximate these hours was:

0 Method A El Method B The individual was a normal shift Operator (i.e., attended requal, no illness or absence > 30 days per year).

The individual was not a normal shift Operator or was assigned offshift during some of this reporting period.

El Method C Other.... attach details.

Method A 1350 hrs/yr X __

years = __

hours Method B (1350 hrs/yr X years) - (160 hrs/mo X mo*) =

hours

• months off shift or other shift absence > 30 days PERFORMED BY:

INDEPENDENTLY REVIEWED BY:

/

Office Assistant or SRO Ops Supvr Date

/

Shift Manager Date FILE DISPOSITION Original:

CDM Cc:

Compliance Shift Manager INFORMATION USE ATTACHMENT 3 Page 1 of 1

S0123-0-Al ATTACHMENT 4 Rev 41 WHAT 1 OCFR50.54(x) states:

10CFR50.54(y) states:

Sa 123-0-Al Page 33 of 34 ENTRY INTO 10CFR50.54(X) OR 10CFR50.54(Y)

(NN 200345543)

"A licensee may take reasonable action that departs from a license condition or a technical specification (contained in a license issued under this part) in an emergency when this action is immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications that can provide adequate or equivalent protection is immediately apparent."

"Licensee action permitted by paragraph (x) of this section shall be approved, as a minimum by a licensed senior operator prior to taking the action."

WHEN Declaration of a 50.54(x) and 50.54(y) is required when a procedure does not provide steps that in the judgment of the Shift Manager are immediately needed to protect the public health and safety and no action consistent with license conditions and technical specifications can provide adequate or equivalent protection is immediately apparent.

ACTIONS Shift Manager perform the following:

1. Formally declare entry into 10CFR50.54(x) and (y) to the Control Room Team.

2. State the specific action being taken.

For example, IF an Emergency Operating Instruction does not include a specific step(s) which the SM determines are necessary to protect the public health and safety, THEN the specific action necessary must be clearly stated.

3. State the specific license condition.

For example, deviation from the Technical Specification that will not be followed.

4. Clearly state the reason(s) for the departure.

For example, Operators will align the Containment Spray Pump to inject into the RCS and spray into Containment in parallel. This action is not specifically stated in the EOls and is needed to prevent over-pressurizing the Containment, which has a known breach to the atmosphere.

5. Determine, to the extent possible, when the license condition or Technical Specification will be restored.

6. Communicate all of the above to the SED (Site Emergency Director) and log the above actions per S0123-VIII-10, Emergency Coordinator Duties.

7. Complete a Notification to the NRC including the above information per S0123-0-A7.

8. Log the above items.

INFORMATION USE ATTACHMENT 4 Page 1 of 1

SO 123-0-Al ATTACHMENT 5 Rev 41 S0123-0-Al Page 34 of 34

SUMMARY

OF CHANGES Author:

CJ Schmitt PAX:

84979 Location:

K40W DesonitijSne ehbaiig Request Revise Control Room Manning to Shift Flynn DNA 2, 14, 31 Manning TYPO Correct typo of unnumbered attachment Flynn DNA 30 INFORMATION USE ATTACHMENT 5 Page 1 of 1