QUALITY-RELATED COMANCHE PEAK STEAM ELECTRIC STATION STATION ADMINISTRATION MANUAL CONTROL OF HIGH RADIATION AREAS PROCEDURE NO.

Transcription

1 COMANCHE PEAK STEAM ELECTRIC STATION STATION ADMINISTRATION MANUAL QUALITY-RELATED CONTROL OF HIGH RADIATION AREAS PROCEDURE NO. STA-660 REVISION NO. 7 SORC MEETING NO.: 957-// DATE: Oe,,i-Z'7-?o' EFFECTIVE DATE: O0-f-/-L PREPARED BY (Print): 51,77" A5 l->dlx / TECHNICAL REVIEW BY (Print): FILL k / ol-e-s APPROVED BY:. PLANT MANAGER EXT: "'' EXT:.2C2f "L DATE: 1-4 -'5

2 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 2 OF PURPOSE The purpose of this procedure is to prescribe the required controls for entry into High Radiation Areas, Locked High Radiation Areas and Very High Radiation Areas at CPSES. 2.0 APPLICABILITY This procedure applies to all personnel requiring entry into High Radiation Areas, Locked High Radiation Areas or Very High Radiation Areas. 3.0 REFERENCES 3.1 STA-657, "ALARA Job Planning/Debriefing" 3.2 CPSES Units 1 & 2 Technical Specification 6.12, (ITS 5.7), "HIGH RADIATION AREA" CFR 20, "Standards for Protection Against Radiation" 3.4 EPP-201, "Assessment of Emergency Action Levels, Emergency Classification and Plan Activation" 3.5 RPI-606, "Radiation Work and General Access Permits" 3.6 RPI-612, "Steam Generator Work Control" 3.7 NRC Regulatory Guide 8.38, "Control of Access to High and Very High Radiation Areas in Nuclear Power Plants" 4.0 DEFINITIONS/ACRONYMS 4.1 Dose Margjn - The remaining allowable total effective dose equivalent an individual may receive during a specified monitoring period. Ak

3 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 3 OF High Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 100 mrem in 1 hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. 4.3 Locked High Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving a dose equivalent in excess of 1000 millirem in 1 hour at 30 centimeters from the radiation source or from any surface that the radiation penetrates. 4.4 RAD Key - Those keys which provide access to areas which are controlled by Radiation Protection. RAD keys shall not be used to control security related area doors. 4.5 Electronic Dosimeter - A radiation monitoring device which continuously integrates the radiation dose rate and alarms when a preset integrated dose or dose rate is received. 4.6 Stay Time - The length of time ar, adividual may be allowed into an area based on radiation levels and remaining dose margin. 4.7 Very High Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in 1 hour at 1 meter from a radiation source or from any surface that the radiation penetrates. 4.8 Whole Body - Defined as 30 centimeters (12 inches) from the source of radiation. For the purposes of external exposure, the whole body includes the head, trunk (including male gonads), arms above the elbow, or legs above the knee. 5.0 RESPONSIBILITIES 5.1 Radiation Protection (RP) is responsible for the following: Performing routine surveys of the plant and identifying High Radiation, Locked High Radiation and Very High Radiation Areas Posting High Radiation Areas, Locked High Radiation Areas and Very High Radiation Areas with warning signs and appropriate barriers.

4 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO.7 PAGE 4 OF Installing a "RAD" lock on Locked High Radiation Area Doors (which are not security related doors) and unique locks on Very High Radiation Area doors and issuing "RAD"' keys Maintaining administrative control of the keys to all Locked High Radiation and Very High Radiation Areas Issuing Radiation Work Permits to control work in High Radiation Areas and Locked High Radiation Areas Notifying the Shift Manager when areas are classified as either Locked High Radiation or Very High Radiation Areas. I! The Radiation Protection Manager is responsible for maintaining this procedure current. 6.0 INSTRUCTIONS A 6.1 High Radiation Areas Each High Radiation Area shall be posted in accordance with approved Radiation Protection procedures/instructions. [C] All entries into High Radiation Areas shall be controlled by a Radiation Work Permit (RWP). [C07652] [C00796] [C06097] All entries shall be provided with or accompanied by one or more of the following: I A radiation monitoring device which continuously indicates the radiation dose rate in the area, or A& WI folt/f

5 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 5 OF Electronic Dosimeter. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them, or An individual qualified in radiation protection procedures equipped with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the RWP During emergency conditions as established in EPP-201, the Emergency Coordinator may approve entry into High Radiation Areas without an RWP if continuous coverage by Radiation Protection personnel is provided Job planning is required prior to starting work.. a High Radiation Area in accordance with STA A list of High Radiation Areas, Locked High Radiation Areas and Very High Radiation Areas should be maintained at Access Control. 6.2 Locked High Radiation Areas NOTE: RAD locks are used for controlling access to Locked High Radiation Areas only. Areas in the containment buildings which are normally Locked High Radiation Areas may have the locks left in place during a shutdown if changing the lock core is impractical (e.g., plant trip with expected shutdown limited and access to the locked areas is minimal) In addition to the High Radiation Area requirements of Section 6.1 of this procedure, entrances to Locked High Radiation Areas shall be locked. If a Locked High Radiation Area has no enclosure which can be locked and no enclosure can be reasonably constructed around it, then that area shall be barricaded, conspicuously posted, and a flashing light activated as a warning device.

6 STON CPSES AAREAS OFMIGHSRATION PROCEDURE NO. REVISIONNNO.7NUAL STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 6 OF When it is determined that an area is a Locked High Radiation Area, Radiation Protection shall notify the Shift Manager. NOTE: When an area controlled by security becomes a Locked High Radiation Area, key control should be sufficient to ensure positive control to the area is maintained. This may be accomplished by use of existing alarms, shift orders or other instructions Upon becoming a Locked High Radiation Area, Radiation Protection should replace the existing lock with a RAD lock, except when the area is controlled by security. (FX ) [C] Continuous Radiation Protection coverage shall be provided at all times when personnel are in the area. Alarming electronic dosimetry or Telemetry may be used in lieu of continuous RP coverage if the dose rates are known and allowed for in the RWP. [C 02302] A Locked High Radiation Area Entry Authorization Form STA should be initiated for each entry into the area. The form may be computer generated Form STA is not required to be initiated for Radiation Protection technicians entering a posted locked high radiation area for purposes of deposting the area after a plant transient or evolution has resulted in reduced radiation levels (i.e., deposting after completion of a resin transfer). Entry shall continue to be subject to the requirements of Section 6.1. Upon entry, with the intent of deposting the area, if any measured dose rates greater than or equal to 1R/hr at 1 foot are encountered, the technician shall immediately exit the area and initiate Form STA for subsequent entries Form STA is not required to be initiated if entries are.into steam generators. Steam Generator entries shall be controlled in accordance with RPI-612, "Steam Generator Work Control." A /20

7 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 7 OF 11 NOTE: Exposure alarm settings or stay times should limit the individual's exposure to 80% of their exposure margin. [C] If alarming electronic dosimetry is used, the dose and dose rate set points shall be noted on Form STA If alarming electronic dosimetry is not used, Radiation Protection shall calculate allowable stay times and note them on Form STA [C 06097] The entry shall be approved by a Radiation Protection Supervisor prior to the entry (may be via telecom) A pre-job briefing, including the individuals entering the area and Radiation Protection personnel, should be held concerning radiation levels, the tasks to be performed and stay times (or alarm settings if alarming electronic dosimeters are used) Only an individual qualified in radiation protection procedures may check out a "RAD" key for the area. After the entry is completed, the RAD key should be returned to the point of issue During emergency conditions, as defined in EPP-201, a RAD key may be obtained from Radiation Protection. t Upon determination of an area becoming a Locked High Radiation Area or after accessing the Locked High Radiation Area, the status of the area should be verified in accordance with Attachment 1. A 6t ((t-lq - I -w

8 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 8 OF Very High Radiation Areas Very High Radiation Areas shall be maintained locked. Custody of keys to doors leading into these areas shall be maintained by the Radiation Protection Manager Keys used in security doors for access to Very High Radiation Areas shall be maintained by Security; however, these keys shall not be issued without the permission of the Radiation Protection Manager Very High Radiation Areas shall be secured with a lockable room or plant area. If necessary, controls may be extended to adjacent areas to include lockable doors and rooms to establish positive access control to prevent access to these areas When it is determined that an area is a Very High Radiation Area, Radiation Protection shall notify the Vice-President of Nuclear Operations Upon becoming a Very High Radiation Area, Radiation Protection shall ensure that unique locks are installed in the door(s) leading into the affected areas. These locks shall be other than those used as "RAD" locks If entry is required into a Very High Radiation Area, then every effort shall be expended to eliminate the Very High Radiation Area, or reduce dose rates in the area prior to entry (e.g., system flushes, engineering controls, use of robotics, or elimination of source) Entry into Very High Radiation Areas will normally not be allowed. If entry is required, then the following steps shall be performed: A specific RWP approved by the Radiation Protection Manager is required for entry into a Very High Radiation Area The ALARA Committee shall review the need and purpose for such an entry and grant approval prior to entry. L&~~V I&k

9 CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 9 OF Radiation Protection shall calculate allowable stay times. These calculations shall be reviewed by the ALARA Committee as part of the approval process Form STA shall be initiated for each individual entering the area An ALARA review shall be performed and results submitted to the ALARA Committee as part of the approval process The entry shall be approved by the Vice President of Nuclear Operations or his designee just prior to entry. If the designee provides approval, the Vice President of Nuclear Operations should be informed of the impending Very High Radiation Area entry A pre-job briefing, including the individuals entering the area, Radiation Protection personnel, the Radiation Protection Manager, and line management shall be conducted. This briefing shall include, as a minimum, radiation levels, stay times, contingency measures and tasks to be performed A direct reading dosimetry system, with remote readout capabilities, shall be utilized for each entry team member A Radiation Protection Technician should accompany the individual to the entrance of the Very High Radiation Area to determine radiation exposure conditions at the time of entry and render assistance, if necessary Direct line of sight shall be maintained with those individuals entering the area and the control point or outside area personnel. If direct line of sight cannot be established, then communication devices shall be utilized by each member of the team entering the area.

10 CPSES STATION ADMINISTRATION M[ANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE10 OF FIGURES None 8.0 ATTACHMENTSIFORMS 8.1 Attachments 8.2 Forms Locked High Radiation Area Verification Form STA-660-1, Locked/Very High Radiation Area Entry Authorization 9.0 RECORDS When completed, the following forms, reports, or other documents generated in response to this procedure shall be dispositioned in accordance with STA-302, "Station Records". 9.1 Form STA-660-1, Locked/Very High Radiation Area Entry Authorization

11 "CPSES STATION ADMINISTRATION MANUAL PROCEDURE NO. STA-660 CONTROL OF HIGH RADIATION AREAS REVISION NO. 7 PAGE 11 OF 11 LOCKED HIGH RADIATION AREA VERIFICATION ATTACHMENT 1 Page 1 of 1 * Radiation survey performed and documented in accordance with plant procedures. * Ensure the lock core has been changed appropriately ('RAD' or 'Very High RAD'). When the area is able to be locked, a Locked high radiation area should be locked with a 'RAD' lock when possible except when positively controlled by security as noted in section 6.2 of this procedure. * Area or room has been verified locked. 0 When area cannot be locked, ensure the area is barricaded, and a flashing light is located in close proximity to the area. * Ensure all entrances have been posted. In lieu of "CAUTION" signs, use 'DANGER" sign for Locked High Radiation Areas, and "GRAVE DANGER" sign for Very High Radiation Areas. 0 Ensure all adjacent areas have been posted as appropriate. An example may be that a Locked High Radiation Area exists in an open area. The appropriate posting may include; Locked High Radiation Area at the source, High Radiation Area several feet from the source, and a Radiation Area at a further distance from the source. * Ensure Shift Operations has been notified. * Ensure the list of Locked High Radiation Areas (Very High Radiation Areas) is updated. * Ensure a log entry is made in the Radiation Protection Shift Log. AL a/-ýll

12 Locked / Very High Radiation Area Entry Authorization Radiation Work Permit #: Location : Entry Date: / / El Electronic OR El Stay times Dosimeters issued Calculated NAME SSN MARGIN EXPECTED EXPECTED DOSE RATE DOSE * STAY TIME* (totem) RATE (mr/hr) DOSE (mr) ALARM (mr/hr) ALARM (mr) (minutes) Alarm Setpoints or stay time comments * NOT to Exceed 80% of Margin "[1 Prejob Briefing performed and documented in accordance with STA-657 "El Telemetric dosimetry issued OR El continuous coverage provided Radiation Protection: Initial Initial Date: / / Entry Authorized By: Entry Approved By: (Very High Rad Area ONLY) RP Supervisor Vice President of Nuclear Operations Date: / / Date: / / STA Page 1 of 1 Revision 6

13 COMANCHE PEAK STEAM ELECTRIC STATION RADIATION PROTECTION INSTRUCTION MANUAL QUALITY RELATED RADIOLOGICAL SURVEILLANCE AND POSTING INSTRUCTION NO. RPI-602 REVISION NO. 19 EFFECTIVE DATE: "h{ Ic [ PREPARED BY (Print): B. Knowles EXT: 5780 TECHNICAL REVIEW BY (Print): John M. Blaikie EXT: 0844 I APPROVED BY: John R. Curtis. 6,.4."LJ- 1 Z~s DATE: 2 / /o, RADIATION PROTECTION MANAGER

14 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 2 OF PURPOSE The purpose of this instruction is to provide the requirements for scheduling, performing, and documenting radiological surveys and to specify radiological posting and labeling requirements. 2.0 APPLICABILITY This instruction applies to personnel involved in scheduling, performing and documenting radiological surveillance and radiological posting activities. 3.0 DEFINITIONS/ACRONYMS 3.1 Airborne Radioactivity Area - A room, enclosure or area in which airborne radioactive materials, composed wholly or partly of licensed material, exist in concentrations in excess of the derived air concentrations (DACs) specified in 1 OCFR20, Appendix B, Table 1, Column 3 or exist to such a degree that an individual present in the area without respiratory protective equipment could exceed, during the hours an individual is present in a week, an intake of 0.6 percent of the annual limit on intake (ALI) or 12 DAC-hours. 3.2 Annual Limit on Intake (ALI) - The derived limit for the amount of radioactive material taken into the body of an adult worker by inhalation or ingestion in a year. ALI is the smaller value of intake of a given radionuclide in a year by the reference man that would result in a committed effective dose equivalent of 5 rems or committed dose equivalent of 50 rems to any individual organ or tissue. 3.3 Barricade - A rope,, ribbon, or other firmly secured, conspicuous obstacle that (by itself or used with physical barriers such as existing walls or hand railings) completely surrounds the area and obstructs inadvertent entry. 3.4 Breathing Zone - The 3' circular region immediately adjacent to a worker's mouth and nostrils while performing a job or task. I

15 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 3 OF Container - As used in this instruction refers to any type of receptacle in which radioactive material is transferred or stored on a temporary or permanent basis. (Examples include: trash bags, boxes, drums, lead shields, etc. which are intended to contain radioactive material.) 3.6 Contamination Area - Any area, accessible to personnel in which loose surface contamination levels are greater than or equal to: 1000 dpm/100 cm' for beta-gamma radiation or 20 dpm/l100 cm2 for alpha radiation. 3.7 Derived Air Concentration (DAC) - The concentration of a given radionuclide in air which, if breathed by the reference man for a working year of 2,000 hours under conditions of light work, results in an intake of 1 ALI. 3.8 Derived Air Concentration-hour (DAC-hr) - the product of the concentration of radioactive material in air (expressed as a fraction or multiple of the derived air concentration for each radionuclide) and the time of exposure to the radionuclide, in hours (2,000 DAC-hours represent one ALl, equivalent to a committed effective dose equivalent of 5 reins [0.05 Sv]). 3.9 DAC Ratio - Ratio of actual airborne radioactivity concentration of a radionuclide to the DAC value for that radionuclide. Expressed as a fraction or percent DAC (i.e.,.25 DAC, 25% DAC) Disintegration per Count - D/C - The efficiency factor derived for lab counters and friskers, usually located on a sticker on the instrument. For beta gamma friskers the D/C value used is 10. (Instrument efficiency is the inverse of "D/C".) 3.11 General Area Dose Rate - The radiation level to which a major portion of the body may be exposed during normal activities, normally at 30 centimeters from sources of radiation Gross Airborne Radioactivity - The particulate or radioiodine concentration calculated by measuring air samples with an instrument other than a multi-channel analyzer.

16 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 4 OF High Radiation Area - Any area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 100 millirem in 1 hour at 30 centimeters from the radiation source, or from any surface that the radiation penetrates Hot Pipe - A length of pipe along which radiation levels (during non-transient conditions) are relatively evenly distributed and the radiation level from a 30 centimeter dose rate are 5 times higher than the average background general area dose rates for the room or area and the pipe's contact dose rate is > 100 mrlhr. (Hot Pipes are not required to be posted when the pipe is located within a posted HRA, LHRA or VHRA) 3.15 Hot Spot - A localized area such as a valve, pipe bend or crud trap, where the radiation levels from a 30 centimeter dose rate are 5 times higher than the average background general area dose rate for the room or area and the suspect hot spot's contact dose rate is > 100 mr/hr. (Hot spots are not required to be posted when the spot is located within a posted HRA, LHRA or VHRA) 3.16 Labeling - As used in this instruction refers to the process of identifying and designating specific items (i.e., containers, tools, sources, waste, etc. which are intended to contain radioactive material). Labels identify radiological conditions Large Area Wipe Survey - A qualitative contamination survey method used to indicate whether or not loose contamination is present on large surfaces Locked High Radiation Area - Any area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 1000 millirem in I hour at 30 centim~eters from the radiation source or from any surface that the radiation penetrates Non-Routine Survey - Radiation, contamination and airborne radioactivity surveys that are performed on an as-required basis on areas or equipment not normally covered by routine surveys. I

17 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 5 OF Posting - As used in this instruction refers to the process of establishing and designating the following areas: Radiation Area, High Radiation Area, Locked High Radiation Area, Very High Radiation Area, Hot Spots, Hot Pipe, Airborne Radioactivity Area, Contamination Area, Radioactive Material(s) Area, and Radiologically Controlled Area Radiation Area - Any area, accessible to individuals, in which radiation levels could result in an individual receiving a deep dose equivalent in excess of 5 mrem in 1 hour at 30 centimeters from the radiation source, or from any surface that the radiation penetrates Radioactive Materials Area - Any area in which licensed radioactive material is used or stored in amounts exceeding 10 times the quantity of such material specified in Appendix C of 10CFR Radiologically Controlled Area - Any area where access is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials Routine Survey - Radiation, contamination and airborne radioactivity surveys that will be routinely performed according to a schedule Very High Radiation Area - An area, accessible to individuals, in which radiation levels could result in an individual receiving an absorbed (dose deep dose equivalent) in excess of 500 rads in one hour at 1 meter from a radiation source, or from any surface that the radiation penetrates.

18 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 6 OF PRECAUTIONS/LIMITATIONS/NOTES 4.1 Precautions When explosive gases are suspected, good judgement and/or explosive gas meters should be used to determine the need for remote sampling methods. Consult, as necessary, with the Operations Shift Manager to determine the content of the system in question Caution should be exercised in collecting, handling and disposal of air samples and smears to prevent possible cross-contamination. 4.2 Limitations Sr-90 and alpha emitter monitoring is not required until analyses of the Reactor Coolant system indicate the presence of either Sr-90 or alpha emitting radionuclides Due to the absorption capabilities of charcoal cartridges, use one that has been recently removed from a sealed or closed package. 4.3 Notes For the purposes of this procedure, "the primary RCA" refers to the areas that are permanently posted and controlled by the main access control point, Room For the pvrposes of this procedure, 30 centimeters is equated to 12 inches Hot spots and hot pipes are not required to be posted when the pipe or spot is located within a posted HRA, LHRA or VHRA.

19 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 7 OF Some areas in the RCA do not require access on a routine or normal basis. These areas may have conditions that make routine radiation or contamination surveys impractical. These conditions may warrant exclusion from routine scheduling and the conditions may include; elevated radiation levels with low or no routine occupancy and confined spaces (other conditions may apply) File cabinets, desks and other storage containers are used inside the RCA for purposes other than radioactive material storage. These containers do not require labeling for radioactive purposes. However, if any of these devices are found to contain radioactive material, the container should be labeled as soon as practical, or the material removed from the container. [C] Portable instrumentation used for survey purposes shall be able to measure the type and magnitude of radiation hazard expected. (C08227) 5.0 PREREQUISITES 5.1 All radiological signs and labels used shall have a yellow background. At least one conventional magenta or purple colored three bladed radiation symbol must appear on each sign in accordance with 10CFR20, Section The color requirement does not apply to inserts. Postings shall be conspicuous. 5.2 Signs, labels, rope, tape or any other materials used to designate radiologically controlled areas should not be used for any other purpose. Such materials should be capable of withstanding the environment they are placed in without being degraded to the point of inadequate control.

20 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 8 OF Prior to posting a new area, ensure all personnel within the area are notified of the posting and/or direct them to leave the area. Note: Containment surveys are not required to be posted at the doorways inside the containment buildings. Other methods of presenting the survey information are acceptable, such as note books containing the surveys at the control points, verbal updates by the technician covering the job, or computer information (written or pictorial) which may provide survey data to the radiation worker. 5.4 Update maps at the entrance to rooms after completion of surveys or changing postings as appropriate, if significant changes in radiological conditions have occurred. As an alternative, during periods of plant transient conditions, direction to contact RP for current radiological conditions may be posted at room entrances with approval from an RP Supervisor. 6.0 INSTRUCTIONS Radiological surveys shall be made to ensure that CPSES is in compliance with 10 CFR 20, Section Surveys should be reasonable under the circumstances to evaluate; "* the magnitude and extent of the radiation levels, "* the concentrations or quantities of radioactive material, and "* the potential hazards. Instrumentation used for radiological surveys shall be calibrated in accordance with station procedures when used for historical documentation or when the survey information will be used to establish dose rates for area posting. I

21 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 9 OF Radiological Surveillance [C] Note: Some areas in the RCA do not require access on a routine or normal basis. These areas may have conditions that make routine radiation or contamination surveys impractical. These conditions may warrant exclusion from routine scheduling and may include; elevated radiation levels with low or no routine occupancy, confined spaces. The frequency of routine surveys shall be based on the potential radiological hazards, probability of change in radiological conditions, and frequency of occupancy of the areas involved. Permanent platforms and temporary scaffolding should be surveyed at the same frequency as routine radiation and contamination surveys for the area in which they are located. If an area is not surveyed routinely, a sign, label, or other marking should be clearly visible to indicate that Radiation Protection must be contacted for current radiological information. (CO1492) I Routine radioactive contamination surveys should be conducted as follows: 1) bi-weekly in Primary Sample Rooms, Chemistry Hot Lab, General Walkways and Change Areas within the RCA. 2) bi-weekly in areas of the RCA where radioactive materials are handled or stored, or where contamination boundaries or postings are located. 3) monthly in lunch/eating areas inside the protected area. 4) monthly in some areas outside the RCA but within the protected area (i.e., offices, shops, and storage areas). These areas should be checked on a rotating basis so all areas are surveyed at least every six months.

22 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 10 OF Routine radiation surveys should be conducted as follows: [C] 1) Surveys for the Non-RCA Trash Trailers located in the Protected Area shall be performed weekly. (C25872) 2) bi-weekly in areas of the RCA where personnel frequently work or enter. 3) monthly - areas in the RCA of non-radiological impact. These areas should be checked on a rotating basis so all areas are surveyed at least every six months. 4) bi-weekly in radioactive material storage areas. Survey pathways within the storage areas and the external perimeters. 5) Quarterly - Areas outside the RCA but within the protected area (i.e., offices, shops and storage areas) Airborne radioactivity surveys shall be conducted in radiologically controlled areas where there is the potential for airborne radioactivity. Evaluation and/or performance of airborne radioactivity surveys should include but not be limited to the following: [C] 0 airborne radioactivity surveys shall be performed during any work or operation known or suspected to cause airborne radio-activity, such as grinding, welding, burning, cutting, hydro lazing, use of compressed air or volatiles on contaminated equipment, or during removal of contaminated insulation (CO 1492) "* during any work or operation that involves the breach of a radioactive system where the potential for airborne radioactivity exists "* inside the containment building during modes 5 and 6 I "* any time respiratory protection devices or alternate tracking methods (DAC-hours) are used to control internal radiation dose I

23 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 11 OF 46 [C] Continuous air monitors shall be utilized in all normally occupied plant areas where the likelihood for airborne radioactivity exists. (CO 1492) Surveys in areas of extreme radiological conditions, where conducting routine survey activities would compromise ALARA principles, should be considered non-routine and be scheduled on an as-needed basis or in accordance with RWP requirements Consideration should be given to increasing the frequency of some routine surveys during outages. (i.e., lunchroom, tool issue areas, containment access area). [C] Digital Radiation Monitoring System radiation and airborne alarms shall be validated (to confirm an alarm). If the alarm is valid (i.e., not related to DRMS computer or instrument malfunctions), surveys shall be performed when practical to identify and isolate the source of abnormal radiation or airborne radioactivity. (C01897, C12619) Scheduling Routine Surveys A Routine Survey Schedule Database should be maintained. This database may be electronic or hardcopy. Survey areas may be added, deleted or rescheduled by an RP Supervisor or a responsible technician based on specific situations, radiological hazards, and/or plant status. I

24 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 12 OF Non-Routine Surveys Note: Normally non-routine surveys are not required in areas identified in Attachment 8. Routine surveys of these areas should be used to provide information to radiation workers. Non-Routine Surveys should be performed as required to ensure adequate knowledge of radiological conditions prior to, during, and/or after any evolution involving exposure or potential exposure to radiological hazards. Document non routine surveys using the RPI survey form. These surveys should be considered valid for two weeks Hot Spot Inspection Verification and Documentation Note: Hot spots and hot pipes are not required to be posted when the pipe or spot is located within a posted HRA, LHRA or VHRA Hot spots should be inspected on a routine basis to ensure that areas are properly posted as necessary. The frequency of these inspections should take into consideration ALARA concerns, location and plant operating modes. However, accessible areas (i.e., outside of containment) should be inspected on a bi-weekly basis, consistent with the survey frequency requirements for these rooms As a minimum, the following should be verified when the inspection of the Hot Spots are performed. "* Hot spot stickers/labels are in good repair (e.g., readable). "* Posting is conspicuous (e.g., not hidden behind piping). "* The general area dose rate has not changed (increased or decreased) to require a room to be reposted. "* The listing is updated as necessary (RPI-602-8). I

25 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 13 OF Information gathered during the hot spot inspection should be used to determine any corrective actions required to reduce the hot spot (e.g., flushing, draining or other maintenance activities) Document the hot spot(s) on RPI by clearly designating the location with the word "HOT SPOT." The abbreviation "HS" may be used if it is defined elsewhere on the RPI form. The component or description should be documented, with a contact and 12" radiation reading. The hot spot contact dose rate should be recorded at the bottom right comer of the RPI form. If multiple hot spots are identified on the RPI form, record the highest hot spot reading at the bottom right comer of the form. See figure 1 as example. H.,-/ I -o I CL- - Sv-- /... A.- CA DOSE RECRIVED PERFORMING THIS SURVEY:.. _ API Figure 1 Example of recording hot spot information. [C] High Radiation Area Inspection (C07652) High, Locked High and Very High Radiation Areas shall be inspected on a routine basis to ensure that areas are properly posted, barricaded or locked as necessary. The frequency of these inspections should take into consideration ALARA concerns, location and plant operating modes. However, accessible areas (i.e., outside of containment) should be inspected on a bi-weekly basis, consistent with the survey frequency requirements for these rooms.

26 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 14 OF As a minimum, the following should be verified when inspections of any HRA, LHRA, VHRA is performed. "* Boundary and posting material is in good repair. "* Posting is conspicuous. * Signs are clearly visible with doors or gates open or closed. * Locks (for LHRA or VHRA) are locked. [C] * If the area is a LHRA, and no locking mechanism is available, the area shall be barricaded, and has a flashing light. (C26935) * The listing is updated as necessary (RPI-602-7) HRA/LHRA postings in the containment buildings are not required to have a weekly verification performed. The verification should be performed as necessary to maintain positive control of these posted areas. Attachment 5 is a sample of a survey data sheet that may be used to verify containment HRA/LHRA postings. 6.2 Survey Performance Perform appropriate surveys and posting using Attachments 1 and 2 as a guideline Daily routine surveys should be performed once each day Weekly/Bi-weekly routine surveys should be performed during the designated scheduled 7 day period Monthly routine surveys should be performed at any time during the scheduled calendar month Quarterly surveys should be performed anytime during the scheduled calendar quarter Non-routine surveys performed in specified routine survey areas during the scheduled period may be used for routine survey purposes.

27 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 15 OF Routine surveys containing appropriate data may be used for non-routine purposes Permanent platforms and temporary scaffolding should be surveyed for radiation and contamination as part of any routine survey Temporary RCAs should be surveyed daily (unless the area is included in the routine survey package, at which time the survey package frequency is sufficient). 6.3 Documenting Surveys [C] Completed surveys shall be used to provide a historical basis for decision making to provide the best protection for radiation workers. When surveys are complete, the originals or copies may be used to communicate adverse radiological conditions. (C2685 1) Complete applicable data for the surveys as appropriate on RPI Detailed instructions are given in Attachment 3 to this procedure Ensure form RPI-602-7, "Current List of High Radiation/Locked High Radiation/Very High Radiation Areas" is updated as necessary to reflect the current status of all appropriate areas Ensure form RPI-602-8, "Current List of Hot Spots" is updated as necessary to reflect the current status of all appropriate areas. 6.4 Radiation Surveys When performing general area surveys attempt to locate sources of radiation When a hotspot is identified, dose rates should be measured and recorded on contact and at 30 cm Contact measurements are those dose rates performed at no more than 1" from the source of radiation.

28 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 16 OF Document radiation surveys on Form RPI Contamination Surveys Frisker probes should be held no more than 1/2" from the sample or surface surveyed, and no more than 1/8" for alpha analysis Smears reading less than 2000 cpm may be analyzed on a lab counter Highly contaminated smears (e.g., greater than 500,000 dpm) should be measured with an ion chamber for beta and gamma radiation. Record the results of these measurements as "mrad/hr" or "Rad/hr." The results should be recorded as noted in Attachment Large area wipe surveys should not cover an area too large for the conditions to be adequately assessed. Several conditions may dictate the maximum area a wipe can be used including floor texture, humidity, or housekeeping. Normally large area wipes are limited to about 200 ft 2 (experience and discretion may be used to determine the maximum area for the size of a large area wipe) While the path covered by a large area wipe survey need not be indicated, a statement should be made in the "Remarks" section describing the survey results to include details as to where (on floor and/or walls) the large area survey was taken. [C] Large area wipe surveys taken inside the RCA require a followup smear survey if the highest reading obtained equals or exceeds 100 cpm above background. Results of the followup survey shall be used to determine other actions (e.g., decon, posting, etc.). (C07667) [C] Large area wipe surveys taken in unrestricted areas (outside of the RCA) which indicate greater than background require a followup smear survey counted on a low background (laboratory) counter. Results of the followup survey shall be used to determine other actions (e.g., decon, posting, etc.). (C07677)

29 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 17 OF When conducting large area or direct frisk surveys, frisk slowly and record results in cpm, or as background (bkgd) if no reading above background is obtained Areas that are carpeted and are surveyed for contamination should be direct frisked Record contamination surveys on Form RPI Airborne Surveys Air sampling should be performed as close as practical to the actual or potential breathing zone of personnel occupying the area. Consideration should be given to special conditions which may affect the sample collection such as ventilation air flow, stratification, and surface contamination When sampling for iodine, both particulate filter and iodine cartridge are required. When sampling for particulate only, and iodine cartridge is required if necessary to seal the sample head in addition to the particulate filter Regulated air samplers may be fitted with a length of tubing between the sample head and the sampler inlet. This will allow placing the sample head at a location remote from the sampler One air sample ID number may be used to identify particulate, iodine and gas samples taken simultaneously Grab Samples NOTE: A continuous air monitor (e.g., AMS-4) may be used in lieu of a grab sample Minimum collection volume is 10 cubic feet or 300 liters Flow rate should be limited to approximately 2 cfm (60 1pm) for charcoal cartridges and 4 cfmn (120 1 pm) for silver zeolite cartridges. I

30 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 18 OF When radioiodine is suspected to be present and accompanied by large quantities of noble gases (e.g., 100 DAC or greater) or in the event of a declared emergency, silver zeolite cartridges should be used if available Record sample information on the Air Sample Data Sticker, or equivalent (attachment 7 is a sample of this sticker) Perform documentation and sample analysis in accordance with Section 6.3 and 6.7 of this procedure Noble Gas Sampling Place lid on Marinelli flask and transport to sample location Remove lid and wave flask vigorously several times through the air Replace lid on flask Transfer the flask, along with a completed Air Sample Data Sticker and "Request for Gamma Analysis" to the Chemistry count room for analysis Document results in accordance with 6.3 of this procedure When the activity for Xe-133 exceeds 7E-5 vci/cc, Noble Gas Skin Dose Calculations should be initiated in accordance with RPI Tritium Sampling is normally performed by the Chemistry section.

31 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 19 OF Sample Analysis/Air Sample Documentation Count the air sample on the appropriate counting instrument, (e.g., low background counter or frisker) Using the appropriate form RPI-602-1, perform a calculation of the air sample results. Normally an RPI form designed specifically for air samples should be used for documentation. However, for some routine functions (e.g., NSSS filter changes or reactor cavity work), RP supervision may approve the use of a common RPI for job coverage as well as air sample results. NOTE: If sample is suspected to contain Noble gas or Radon sample may be recounted after 30 min. This does not relieve the requirements for posting during this decay period If the calculated air sample activity for particulate is greater than 2.5 E-9 jici/cc, send the sample with Air Sample Data Sticker and completed "Request for Gamma Analysis" to the Chemistry count room If the charcoal cartridge is counted with a frisker probe and the counts are greater than background, the sample with Air Sample Data Sticker and completed "Request for Gamma Analysis" should be delivered to the Chemistry count room. NOTE: When air sample media are taken to Chemistry for analysis, the MCA results take precedence over any manually calculated values from RPI If results from Chemistry are greater than.25 DAC: I * post in accordance with Attachment 1 * determine airborne exposure tracking needs in accordance with RPI determine if respiratory equipment is justified by completing form RPI (if form has not already been completed for the job, or if the DAC concentration is greater than the completed form)

32 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 20 OF When using a continuous air monitor (e.g., AMS-4) during job coverage the results of the monitor should be documented on the appropriate RPI Negative results should be documented with a statement similar to "No AMS-4 alarms during system breach of system XYZ." When air sample results indicate DAC values greater than 25 percent, then results should be documented on an RPI form designed specifically for air samples. This will ensure the data for the air samples can be easily retrieved. 6.8 Sr-90 and Alpha Emitters Sampling NOTE: The following monitoring program is not required until analysis for the Reactor Coolant System indicate the presence of either Sr-90 or alpha emitting radionuclides On a quarterly basis or more frequently as necessary, obtain smears in a contaminated area Concurrent with the smears, obtain air samples where there is a likelihood for a measurable airborne concentration, preferably in the same area as where the smears were taken Analyze the samples for Sr-90 and for long lived alpha emitting radionuclides Use the results of these samples as an aid to determining the correct DAC values in accordance with 1 OCFR20, Appendix B, Table 1, Column 3.

33 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 21 OF Alternate Radiation Monitoring of Containment Technical Specification requires that an alternate method of ascertaining Containment area radiation levels be initiated when both Containment High Range Radiation Monitors are inoperable with the reactor in Modes 1, 2 or 3. Guidelines for Alternate Containment Monitoring are in Attachment 4. HRRM Unit Channel# Monitor ID Unit 1 CTE RE-6290A CTW RE-6290B Unit 2 CTE-216 2RE-6290A CTW-217 2RE-6290B 6.10 Posting / Material Storage in the Containment Buildings STA-620 provides instructions to ensure the Technical Specification requirements for loose debris in the containment buildings are met. The following allowances are provided for leaving radiation protection related items in the containment building during modes 1 through 4. These allowances are based on the results of Technical Evaluation TE Radiological posting signs (3 or 5 pocket type) may be left in the containment building, provided the signs are firmly attached using tie wraps through the grommets. The sign may also be firmly attached to a rad rope if the rope is firmly attached as noted in Tape should not be used to attach signs Rad rope may be left in the containment building, provided the rope is firmly attached using tie wraps at both ends. Tape should not be used to attach rope. The rope should not be placed on any equipment that may prevent actuation or operation of the equipment.

34 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 22 OF Electric flashing lights and associated equipment should be secured with stainless steel tie wraps Stanchions used for radiological control purposes may be left in the containment building, provided the stanchion is securely fastened in accordance with STA Normally all consumable items such as bags, wipes, maps, etc., should be removed from the containment building prior to the completion of the containment entry. Posting materials should be properly secured to meet the intent of STA-620, STA-661, and STA Verifications should be performed as specified by a Radiation Protection Supervisor to ensure all posting material in the containment building meet this criteria. The verification should be documented on RPI (Attachment 5 is a sample of the verification) Steam Generator Surveys [C] Steam Generator surveys shall be performed in accordance with this instruction to ensure proper posting and control is maintained. Additional information about steam generator surveys may be found in RPI-609, "EPRI - Westinghouse Radiation Monitoring Program," and RPI-612, "Steam Generator Work Control." (C2691 1) 6.12 Records When completed, the following forms, reports or other documents generated in response to this instruction shall be dispositioned in accordance with STA-302, "Station Records" RPI-602-1, "CPSES Survey Data Sheet." RPI-602-7, "Current List of High Radiation, Locked High Radiation/Very High Radiation Areas." RPI-602-8, "Current List of Hot Spots."

35 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 23 OF Copies of all records sent to Document Control should be filed in the RP office and retained as long as determined necessary by RP supervision. 7.0 REFERENCES OCFR20, "Standards for Protection Against Radiation" 7.2 STA-302, "Station Records" 7.3 STA-620, "Containment Entry" 7.4 STA-650, "General Health Physics Plan" 7.5 STA-652, "Radioactive Material Control" 7.6 STA-660, "Control of High Radiation Areas" 7.7 STA-66 1, "Non-Plant Equipment Storage and Use Inside Seismic Category I Structures" 7.8 STA-729, "Control of Transient Combustibles, Ignition Sources and Fire Watches" 7.9 RPI-213, "Survey and Release of Material and Personnel" 7.10 RPI-230, "Radioactive Material Shipments" 7.11 RPI-240, "Shipment of Radioactive Waste to Waste Processors" 7.12 RPI-254, "Shipment of Radioactive Waste to Disposal Facilities" 7.13 RPI-506, "Calculation and Tracking of Personnel Exposures to Airborne Radioactive Material" 7.14 RPI-606, "Radiation Work and General Access Permits" 7.15 RPI-609, "EPRI - Westinghouse Radiation Monitoring Program,"

36 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 24 OF RPI-612, "Steam Generator Work Control." 7.17 RPI-801, "Operation of Portable Survey Instruments" 7.18 RPI-802, "Performance of Source Checks" 7.19 RPI-9 11, "Use of Lapel Air Sampler" 7.20 NUREG NE-27698, "Dose Rate Ratios For Alternate Radiation Monitoring" 7.22 NUREG/CR Comanche Peak Steam Electric Station Technical Specifications 7.24 Technical Evaluation Technical Evaluation ATTACHMENTS/FORMS 8.1 Attachments Attachment 1, Posting Requirements Attachment 2, Posting Guidelines Attachment 3, CPSES Survey Data Sheet Instructions Attachment 4, Guidelines for Alternate Containment Monitoring Attachment 5, Containment HRA!LHRA Verification Survey (sample) Attachment 6, Hot Spot Listing (sample)

37 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 25 OF Attachment 7, Air Sample Sticker (sample) Forms Attachment 8, Non-Radiological Impact Areas. I RPI-602-1, "CPSES Survey Data Sheet" RPI-602-7, "Current List of High Radiation/Locked High Radiation/Very High Radiation Areas" RPI-602-8, "Current List of Hot Spots"

38 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 26 OF 46 ATTACHMENT 1 PAGE 1 OF 6 [C] POSTING REOUIREMENTS (CO1492, 05756, and 26739) NOTE: For purposes of posting, individual dose rates (i.e., gamma and neutron) should be considered cumulative. For example an area with a gamma dose rate of 800 mr/hr and a neutron dose rate of 250 mrem/hr should be posted as a Locked High Radiation Area. Beta dose is not a deep dose equivalent and should not be considered for posting purposes. IF dose rates to the major portion of the whole body exceed: 5 millirem/hr inside the primary RCA or 2 millirem in any one hour outside the primary RCA and authorized by RP supervision or **0.5 millirem/hr in an unrestricted area continuously. THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION RADIATION AREA ** True "members of the public" have restricted access within the Owner Controlled Areas. Unmonitored (non-radiation workers) employees are the most likely individuals to access radiation areas outside the primary RCA. I Security patrols outside the Protected Area every 24-hrs for "suspicious" behavior. Assuming 2000 hours/year for employees and 1/16 Occupancy factor for "occasional" (from NCRP Report #49, Appendix C, Table 4) results in 0.8 mrem per hour. To ensure compliance, boundaries/barriers to prevent unmonitored personnel entry into Radiation Areas should be posted at 0.5 mrem per hour.

39 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 27 OF 46 ATTACHMENT 1 PAGE 2 OF 6 POSTING REQUIREMENTS NOTE: Signs and postings with the word "DANGER" may be used for HIGH RADIATION AREAS in the containment buildings during mode changes, such as after a refuel outage when expecting to startup. IF dose rates to the major portion of the body exceed 100 millirem per hour, THEN: The area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION (or DANGER) HIGH RADIATION AREA A High Radiation Area shall be controlled in accordance with STA-660, "Control of High Radiation Areas" and as a minimum shall have entry controlled by a GAP or RWP and shall be barricaded and conspicuously posted. IF dose rates to the major portion of the body exceed 1000 millirem per hour, THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION (or DANGER) LOCKED HIGH RADIATION AREA A Locked High Radiation Area shall be controlled in accordance with STA-660, and as a minimum shall be posted and locked, or if it cannot be locked, it shall be barricaded and have a flashing light as a warning device. Continuous RP coverage is required for entry. A barricade and flashing light should not be used if the area can be locked or when a lockable enclosure can reasonably be installed.

40 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 28 OF 46 ATTACHMENT 1 PAGE 3 OF 6 POSTING REQUIREMENTS IF dose rates to the major portion of the whole body are such that an individual could receive an absorbed dose in excess of 500 rads in I hour at 1 meter from the source, THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: GRAVE DANGER VERY HIGH RADIATION AREA Very High Radiation Areas shall be controlled in accordance with STA-660. IF AND AND the suspect hot spot's 30 centimeter dose rate exceeds five (5) times the average general area background dose rate, the contact dose rate is 100 millirem per hour or greater, the area is not a posted HRA, LHRA or VHRA, THEN: the spot should be posted with a label worded: HOT SPOT IF AND AND a run of piping exhibits a dose rate at 30 centimeters that exceeds five (5) times the average general area background dose rate for the room or area, the contact dose rates on the piping are 100 millirem per hour or greater, the area is not a'posted HRA, LHRA or VHRA, THEN: the pipe should be posted (approximately every 6') with labels worded: HOT PIPE I

41 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 29 OF 46 ATTACHMENT 1 PAGE 4 OF 6 POSTING REQUIREMENTS II. IF OR smearable contamination levels exceed 1000 dpm/100 cm 2 beta-gamma radiation, smearable contamination levels exceed 20 dpm/100 cm 2 alpha radiation, THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION CONTAMINATION AREA III. IF airborne radioactivity levels exceed 25% of the amounts specified in 1OCFR20, Appendix "B", Table 1, Column 3 THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION AIRBORNE RADIOACTIVITY AREA

42 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 30 OF 46 ATTACHMENT 1 PAGE 5 OF 6 POSTING REQUIREMENTS IV. RADIOACTIVE MATERIAL NOTE: Signs and labels that warn of existing radiological conditions shall be removed or updated when it has been determined that conditions have changed. IF survey results or source receipt papers indicate Radioactive Material in any amount greater than 10 times the quantity of such material specified in 1 OCFR20, Appendix "ICtc, THEN: the area shall be conspicuously posted with a sign or signs bearing the radiation caution symbol and the words: CAUTION RADIOACTIVE MATERIAL

43 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 31 OF 46 ATTACHMENT 1 PAGE 6 OF 6 POSTING REQUIREMENTS NOTE: 2mR/hr contact reading is based on CPSES DAW nuclide mix for 10CFR20 Appendix C values. (TE ) Containers not intended to store radioactive material are not required to be labeled. However, if these containers are found to have radioactive material stored, the material should be removed as soon as practical or the container should be posted or labeled appropriately. If the radioactive material contact reading is greater than 2 mr/hr, a SmartForm should be initiated in accordance with STA IF survey results or source receipt papers indicate a container of licensed Radioactive Material, THEN: the container shall be posted with a durable, clearly visible label identifying the radioactive contents, bearing the radiation caution symbol and the words: CAUTION RADIOACTIVE MATERIAL The label shall also provide sufficient information to permit individuals handling or using the containers, or working in the vicinity thereof, to take precautions to avoid or minimize exposures.

44 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 32 OF 46 ATTACHMENT 2 PAGE 1 OF 4 POSTING GUIDELINES Overview The intent of posting is to alert personnel of the presence, magnitude and boundaries associated with the various radiological conditions encountered in and around the plant. Posting activities provide personnel with sufficient information to minimize radiation exposure. The regulations specified in 1 OCFR20, Section and do not provide implementing details such as whether a room or building containing a radiation area may be posted at the entrance or every discrete radiation area must be posted. Although the regulations refer to "each" area, this does not preclude the posting of a broader area, particularly when it may be advantageous to utilize existing physical structures and boundaries. Each case should be evaluated to assure that posting practices do not detract from the intent to alert personnel to radiological conditions by desensitizing personnel through over posting. Posting activities are performed as an integral part of radiological surveillance. All routine and non-routine surveys should be evaluated against the posting requirements specified in this attachment. Posting verification is accomplished by the use of updated maps at the entrance to rooms. I

45 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 33 OF 46 ATTACHMENT 2 PAGE 2 OF 4 POSTING GUIDELINES Areas with Frequently Changing Dose Rates Areas with frequently changing dose rates should be posted and controlled at the most restrictive level, based upon anticipated fluctuations in dose rates. For example, the volume control tank room may be posted and controlled as a high radiation area while at power, even though general area dose rates within the room may be less than 100 mrlhr for extended periods of time. Areas not frequently occupied and which are not surveyed on a routine basis (e.g., spent resin storage tank room and pipe tunnels) may be posted to address expected radiation levels with requirement to contact RP prior to entry. Posting Signs Most of the posting signs used at CPSES are equipped with three or more pockets for placement of sign inserts. These inserts are printed with a variety of posting designations and requirements for entry. Requirement inserts may be used if determined to be necessary provided that all signs for a particular posting are the same. All posting signs for an individual area should have identical inserts. The outside surface of posting signs should not be written on or marked. Use blank inserts to add necessary information. Multiple Postings When more than one type of radiological condition exists in an area, multiple posting is necessary. This may be accomplished by placing posting designation inserts which describe all radiological conditions in the same area.

46 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 34 OF 46 ATTACHMENT 2 PAGE 3 OF 4 POSTING GUIDELINES Barrier Material Radiological barrier rope should be used for permanent postings. Radiological barrier ribbon should be used for temporary postings but may be used for permanent postings where the use of rope is not practical. Radiological barrier tape may be used on flat surfaces such as floors or work tables where the use of rope or ribbon is not practical. Tape may also be used in conjunction with rope or ribbon to help delineate boundaries or as a labeling device to draw attention to radiological postings. Labeling Containers Radioactive Material contained in items such as laundry bags, bags of waste or tools shall be surveyed, labeled and handled in accordance with RPI-213, "Survey and Release of Material." Radwaste containers shall be labeled in accordance with the appropriate labeling instructions identified in RPI-230, "Radioactive Material Shipments", RPI-240, "Shipment of Radioactive Waste to Waste Processors" and RPI-254, "Shipment of Radioactive Waste to Disposal Facilities".

47 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 35 OF 46 ATTACHMENT 2 PAGE 4 OF 4 POSTING GUIDELINES The following conditions identify when labeling or posting is NOT required: Containers holding licensed material in quantities less than the quantities listed in 1 OCFR20, Appendix C. Containers holding licensed material in concentrations less than those specified in 1 OCFR20, Appendix B, Table 3. Containers attended by an individual who takes the precautions necessary to prevent the exposure of individuals in excess of the limits established by 1 OCFR20. Containers in transport, packaged and labeled in accordance with the regulations of the Department of Transportation. Containers that are accessible only to individuals authorized to handle, use, or work in the vicinity of the containers provided the contents are identified to these individuals by a readily available written record. Installed manufacturing or process equipment such as reactor components, piping and tanks. Areas or rooms containing radioactive materials for periods of less than 8 hours, if each of the following conditions are met: (1) The materials are constantly attended during these periods by an individual who takes the precautions necessary to prevent the exposure of individuals to radiation or radioactive materials in excess of the limits established in 1OCFR20. (2) The area or room is subject to Radiation Protection control.

48 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 36 OF 46 ATTACHMENT 3 PAGE 1 OF 5 CPSES SURVEY DATA SHEET INSTRUCTIONS Instructions for recording data on RPI "CPSES Survey Data Sheet". A. Survey Documentation Section 1. Survey #: Record survey number. 2. RWP/GAP#: In documenting radiological surveys the only surveys needing a RWP/GAP number are those that are "SPECIAL" (if the survey is a "ROUTINE", "N/A" the RWP/GAP # box). Acceptable format for the RWP/GAP number is "Year-RWP/GAP number T - Task number". The Year should be a four digit number, i.e should be "2001" and the RWP/GAP number should be a four digit number, i.e. GAP number 6 (used for Fuel activities) should be "0006". This should be followed by "T", representing "Task" followed by the task number. example: T1 3. Routine/Special: Mark as appropriate. 4. Rx. Power: Record as appropriate. 5. Instrument/ID#/Cal Due Date: Record the model of instrument used (e.g., RSO-5)/ the instrument ID number/ and the calibration due date (the date format is "month / day / year"). 6. Posting Abbreviations: Indicate the appropriate abbreviation on the survey map as to how the area is posted. If a needed abbreviation is not defined, then record the definition in the "Remarks" section. I I

49 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 37 OF 46 ATTACHMENT 3 PAGE 2 OF 5 PSES SURVEY DATA SHEET INSTRUCTIONS 7. Surveyed By: PRINTED NAME of the technician performing the survey, initialed by the same: followed by the date and time of the survey. 8. Reviewed By: Signature of the Technician (normally a Lead Technician) responsible for the review of that survey. (This may also be the RP supervisor responsible for those survey results.) 9. Remarks: Record any pertinent remarks here. This should include the location of the survey if the survey is not on a preprinted form. B. Survey Data Section 1. Survey Maps a. Radiation and contamination surveys are usually recorded on survey maps consisting of floor plans, diagrams of equipment, or other drawings used to show locations of measurements taken during surveys. b. Survey maps are normally pre-printed below the documentation section of the data sheet but, (1) Maps may be hand drawn on blank data sheets. (2) Maps may be on a separate sheet attached to the data sheet. (3) Maps are not required if all survey data and information is recorded accurately on a data sheet. I

50 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 38 OF 46 ATTACHMENT 3 PAGE 3 OF 5 CPSES SURVEY DATA SHEET INSTRUCTIONS 2. Recording radiation dose rates on survey maps a. Record dose rate at location of reading on map. (1) It is not necessary to include dose rate units unless in units other than millirem per hour or millirad per hour. NOTE: Use the symbol "R" to indicate rem or rad. (2) Record gamma dose rates with no suffix (i.e., 200 mr/hr gamma recorded as "200"). NOTE: Beta field dose rates are measured and recorded by using an instrument. Correction factors are not used to record the beta dose rates. ([open window reading] - [closed window reading])= beta dose rate (3) Record beta dose rates with suffix "P3" (i.e., 200 mrad/hr beta recorded as "20013"). NOTE: Beta field dose rates should include distance from source (i.e., 12"). (4) Record neutron dose rates with suffix "N" (i.e., 200 mrem/hr neutron recorded as "200N").

51 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 39 OF 46 ATTACHMENT 3 PAGE 4 OF 5 CPSES SURVEY DATA SHEET INSTRUCTIONS b. Indicate contact readings by placing an asterisk (*) at the location of the measurement on the map and recording the dose rate next to it. Contact readings should be accompanied by a 30 cm reading. Format should follow example below: "*220/25 (i.e., contact dose rate over the 30 cm dose rate separated by a diagonal line) c. If instrument reading is less than the lowest increment of the lowest scale, record dose rate as less than that lowest increment. 3. Recording contamination surveys on survey maps a. Record disc smears by circling the sequential number at the smear location. b. Large area wipes need not be shown on the survey map, but a follow-up disc smear survey is required if the highest reading obtained from any large area wipes equals or exceeds 100 cpm above background. c. Record direct scan by writing net count rate at location of measurement. (1) Include units of ncpm (i.e., 200 ncpm). (2) Net readings indistinguishable from background levels should be recorded as "bkgd".

52 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 40 OF 46 ATTACHMENT 3 PAGE 5 OF 5 CPSES SURVEY DATA SHEET INSTRUCTIONS d. Record contamination results in the margin of pre-printed maps, or on attached sheets. (1) for smears record number and activity the letter "K" may be used to denote multiples of 1000 ex = 3K Smears reading less than 1000 dpmn/ 100 cm 2 may be recorded as "<1K". NOTE: Receipt surveys require alpha contamination surveys. Alpha contamination surveys in plant will not be required until the Chemistry department has informed Radiation Protection that chemistry analysis has confirmed the presence of alpha emitting isotopes. Alpha contamination survey results should be designated by suffix "-". I 3. d. (2) for large area wipe surveys taken inside the RCA reading less than 100 net cpm, a statement should be made in the "Remarks" section of Form RPI to indicate such findings (e.g., "All large area wipe surveys were <I 00 net cpm."). If large area wipe surveys taken in unrestricted areas indicate background, record as "bkgd". 4. Other symbols a. Indicate posting boundary locations by drawing lines intersected by "X's". I

53 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 41 OF 46 ATTACHMENT 4 PAGE 1 OF 2 GUIDELINES FOR ALTERNATE CONTAINMENT MONITORING 1. Set up a Ludlum Model 300 or equivalent to measure contact dose rates within three feet of the centerline of the Containment personnel air lock in room 95 or within three feet (horizontally or vertically) of the point on the wall of room 95, as shown on Figure lb. 2. As directed by the Shift Manager, record the Model 300 reading on the RP shift log. Multiply the dose rate measurement by the appropriate factor from Table 1 and record as in Containment dose rate. 3. Provide copies of the survey data sheet to the Shift Manager. Time After Reactor At Airlock Surface At Wall of Shutdown (hr) Room

54 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 42 OF 46 ATTACHMENT 4 PAGE 2 OF 2 GUIDELINES FOR ALTERNATE CONTAINMENT MONITORING NORTH WALL SAFEGUARDS BLDG ROOM 2-95 SOUTH WALL, SAFEGUARDS BLDG ROOM 1-95, TOP VIEW I 6' X 6' SQUARE I MIDPOINT I i kl i MIDPOINT CONTAINMENT PERSONNEL HATCH SOUTH WALL, SAFEGUARDS BLDG, ROOM 95, FRONT VIEW =iiniss". aa lla ll a a la I ll a Is II. II1 I 6' X 6' SQUARE I

55 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 43 OF 46 ATTACHMENT 5 PAGE 1 OF 1 CONTAINMENT HRA/LHRA VERIFICATION SURVEY (SAMPLE) Containment HIGH RAD AREA Posting / Inspection SURVEY# RWP/GAP# 0 ROUTINE 0 SPECIAL RX POWER1 2 INSTRUMENT / ID# / CAL DUE DATE SURVEYED BY: DATE I TIME REVIEWED BY: DATE POSTING ABBREVIATIONS: RA - RADIATION AREA HRA - HIGH RADIATION AREA LHRA - LOCKED HIGH RADIATION AREA REMARKS: SOP - STEP OFF PAD IC - INTERNALLY CONTAMINATED RMA - RADIOACTIVE MATERIALS AREA AA - AIRBORNE AREA CA - CONTAMINATED AREA CRPPE - CONTACT RP PRIOR TO ENTRY Inspection / Verification for Radiological Posting inside containment during Modes I through 4 LOCATION POSTED AS BOUNDARY VERIFIED DOSE RATE BY: mrlhr 1 st 2nd 808' North of Trash Racks (sign and rope secured with blue tie-wraps) HRA/CRPPE 808' South of Trash Racks (sign and rope secured with blue tie-wraps) HRA/CRPPE 808' RCDT and RCDT Pump Room (sign on gate secured with blue tie-wraps) HRA/CRPPE 819' Let down valve room (sign on gate secured with blue tie-wraps) HRA/CRPPE 822' PRT Room (sign on gate secured with blue tie-wraps) HRA/CRPPE 842' Pressurizer Room (sign on gate secured with blue tie-wraps) HRA/CRPPE 860' Pressurizer Room (sign on gate secured with blue tie-wraps) HRA/CRPPE 905' Pressurizer Room (sign on gate secured with blue tie-wraps) HRA/CRPPE 905' Pressurizer Spray Valve Room (sign on gate secured with blue tie-wraps) HRA/CRPPE No tape has been used for any permanent Posting. (Use Blue tie-wraps for signs and stainless tie-wraps for lights). Both ends of the rope are tie-wrapped when used. All posting materials are secured in place in accordance with STA-661, STA-620, and STA-729

56 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 44 OF 46 ATTACHMENT 6 PAGE I OF I HOT SPOT LISTING (SAMPLE) CURRENT LIST OF HOT SPOTS HOT SPOT LISTING UPDATED: 12/29/98 BLDG ELEV ROOM COMP DESCRIPTION DOSE RATE SURVEY NUMBER AUX 832 X C EAST CORRIDOR 1500/ AUX 842 X-230 XBR-0007 SOUTH CORRIDOR 1200/ SG RH-0008 NORTH EAST CORNER 150/ SG RH-0004 "'B" RHR HX DRAIN LINE 700/ SG RH-0008 SOUTHEAST CORNER 350/ SG RH-0004 "B" RHR HX DRAIN LINE 900/ Reviewed by: kw S Date 12/31/98 (RP Supervisor)

57 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 45 OF 46 ATTACHMENT 7 AIR SAMPLE STICKER (SAMPLE) Air Sample Start Time Flow Rate AIR SAMPLE DATA ID# Start Date Stop Time / Date CFM LPM (circle one) Total Time X Average flow X Correctic )n factor = Total Volume Ii_ I1,pm I L 1000 cfm I"" Started By Task Description Stopped By Location RWP/Task Sampler Mod# / HP# / Cal Due

58 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-602 RADIOLOGICAL SURVEILLANCE AND POSTING REVISION NO. 19 PAGE 46 OF 46 ATTACHMENT 8 NON-RADIOLOGICAL IMPACT AREAS ROOM Electrical Equipment Electrical Equipment Diesel Diesel Diesel Diesel HV Equipment HV Equipment Electrical Safeguards Electrical Safeguards Electrical Equipment & Fuel Oil Day Tank Electrical Equipment & Fuel Oil Day Tank Boron Injection Tank Boron Injection Tank Feed Penetration Areas Feed Penetration Areas Electrical Equipment Area Electrical Equipment Area Mechanical Equipment Area Mechanical Equipment Area Main Steam Penetration Area Main Steam Penetration Area CCW Pump CCW Pump ROOM NUMBER A 2-85A A-D 2-99A-D O0A-H 2-10OA-H / /2-109 X-197, X-198 X-204, X205 I

59 SURVEY* RWP / GAP* E ROUTINE ESPECIALP RX. POWERI l INSTRUMENT / ID# / CAL DUE DATE SURVEYED BY: DATEITIME REVIEWED BY: DATE POSTING ABBREVIATIONS: REMARKS: RPI R-8

60 CURRENT LIST OF NIGH RADIATION AREAS HIGH RADIATION AREA UPDATED: PAGE OF BLDG ELEV ROOM DESCRIPTION j t O LOCKED HIGH RADIATION AREA I I VERY HIGH RADIATION AREA UPDATED BY: RP LEAD TECH / DATE REVIEWED BY: RP SUPERVISOR / I DATE RPI PAGE 1 OF 1 REV. 2

61 CURRENT LIST OF HOT SPOTS HOT SPOT LISTING BLDG ELEV ROOM J UPDATED: COMP DESCRIPTION DOSE SURVEY RATE NUMBER I I I I I I I I I I I I I I I Reviewed by:- (RP Supervisor) Date RPI Page 1 of 1 Rev. 1 I V.

62 COMANCHE PEAK STEAM ELECTRIC STATION RADIATION PROTECTION INSTRUCTION MANUAL QUALITY-RELATED RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS INSTRUCTION NO. RPI-611 REVISION NO. 3 EFFECTIVE DATE: I 1:z)v? PREPARED BY (Print):./1.eI- / TECHNICAL REVIEW BY (Print): X92,IX/77.A, APPROVED BY: fll a~., S6A~l t~eý #ADIATION PROTECTION MANAGER I EXT: 5 9'-z EXT:,,/. DATE: 0 1it' W

63 CPSES RADIATION PROTECTION INSTRUCTION MANUAL 1.0 PURPOSE INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 2 OF 12 The purpose of this instruction is to establish the necessary radiological controls to safely conduct diving operations in the spent fuel pools, reactor cavities, fuel transfer canals, or other areas of radiological concern. The radiological controls are in addition to the requirements of STA APPLICABILIY This instruction applies to Radiation Protection personnel involved in diving operations. 3.0 DEFINITIONS/ACRONYMS 3.1 Diver - a person working under water using surface supplied breathing air or an apparatus which supplies compressed breathing air. 3.2 Diver Tender - a person who assists the diver and tends the tether line and breathing air hoses. 4.0 PRECAUTIONS/LIMITATIONS/NOTES 4.1 Precautions The locations of spent fuel assemblies or irradiated objects should NOT be changed while the diver is in the water. Contact Reactor Engineering to verify pool configuration and planned evolutions which may impact the dive TLD's used for underwater purposes should be placed in water tight wrap to ensure no water penetration Electronic alarming dosimeters should be bagged or wrapped to prevent water damage Caution should be taken to ensure that bubbles do not obscure vision.

64 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI Lmiions RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 3 OF Radiation Protection Personnel assigned to duties associated with the dive at the dive scene should not conduct shift turnover while the diver is in the water Each diver should be equipped with a radiation detector which is remotely read out at the surface and continually monitored by RP personnel RP monitoring personnel should be aware of allowable stay times, available dose margins, stop-work parameters and other pertinent radiological conditions/hazards, both actual and potential, and should ensure these topics are addressed in pre-job briefings Margins, Stay Times, Dose Rates and Dose Alarms should be recorded in the RWP Implementation Sheet, RPI-606-3, in accordance with STA Notes When practicable, physical barriers (e.g.: stainless steel panels, or other approved materials) should be provided to prevent a diver from accessing spent fuel elements or other high radiation items or areas. A "man basket" may be used to confine the diver to a specific work area, in lieu of barrier Colored warning barrier fence material is prohibited from use in the spent fuel pool or other fluid systems in the RCA Each diver should be equipped with a safety line and continuous voice communication with surface personnel providing support for the dive in accordance with STA Emergency guidelines for diver rescue should be provided, reviewed and understood by individuals involved in the diving operation in accordance with STA-682.

65 CPSES RADIATION PROTECTION INSTRUCTION MANUAL *1 r INSTRUCTION NO. RPI-61.I RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 4 OF Divers should be equipped with a calibrated electronic alarming dosimeter that will function and provide an alarm underwater. This dosimeter should be checked for operability each day before diving operations begin Foreign Material Exclusion (FME) controls are required in accordance with STA perequisites 5.1 A specific radiation work permit (RWP) should be issued to cover a specific evolution of dives for a given area. 5.2 Continuous RP coverage should be provided at all times while a dive is in progress. Radiation Protection personnel providing continuous coverage should be appropriately qualified (e.g., mockup training, pre-job training sessions, radiological implications and other applicable radiological facets of the job). 5.3 Radiation Protection management should provide clear expectations on RP stop-work authority to all personnel involved with dive. 5.4 A Pre-Job briefing should be conducted each shift. 5.5 A sample of the pool water shall be obtained and analyzed for tritium and isotopic activity. 5.6 Divers shall provide baseline and final tritium urinalysis samples when diving in pools of water with tritium levels greater than 0.01 tci/cc, in accordance with RPI-500, "Bioassay Program." 5.7 Supplied Air should be tested and verified to meet the requirements of RPI-909, "Testing of Breathing Air Systems," or the air shall be certified by the vendor that is supplying it.

66 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 5 OF INSTRUCTION 6.1 Radiation Surveys Prior to diving operations a survey of the spent fuel pool, transfer canal or reactor cavity where the dive is to take place should be completed using two (2) independent instruments. The dose rates obtained from these surveys should be compared and discrepancies greater than 40% should be investigated Locations of spent fuel assemblies and irradiated objects should be obtained from Reactor Engineering, verified and noted on the survey A comprehensive survey of the diver's path and work area should be completed shiftly during diving operations Continuous monitoring of the diver's work area may be accomplished by attaching an underwater detector to the diver or allowing the diver to carry an underwater detector to the work area. The detector readout should be located on the surface, and the diver should be apprised of the dose rates. 6.2 Pre-Dive Considerations The Pre-Dive Operations Check List (Attachment 1) shall be completed prior to activating the RWP and shiftly for long-term dive evolutions A contamination control area should be established where the diver will exit the water Demineralized water hose should be available to rinse the diver upon exit from the water. Refer to RPI-400 for requirements on using demineralized water Flotation devices should be located near the diving area in case of an emergency.

67 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 6 OF Monitoring Dives The well being and safety of the diver are of prime concern during diving operations. To ensure the safety of the diver while the dive is in progress, the RP Technician responsible for job coverage at the dive site should perform the following, using the instructions in Attachment 2: 1) Maintain communications with the diver through the use of a head set or through the diver tender. 2) Maintain unobstructed visual contact with the diver. A viewing box is preferred, with no ripples, bubbles or other anomalies to prevent a clear view of the diver. Underwater cameras should also be used, where appropriate. 3) Monitor the dose rates at the work area and advise the diver of any significant changes Abort the dive and notify RP supervision if any of the conditions of paragraph 4, Part C of Attachment 2 are observed. 6.4 Completion of Dive The diver's suit and equipment should be rinsed off with demineralized water and dried as much as possible after exiting the water The inside of the diver's suit should be surveyed for contamination All dosimetry should be returned to the dosimetry section if diving operations are complete The post dive worksheet should be completed (Attachment 3) The FME log should be balanced in accordance with STA-625.

68 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 7 OF REFERENCES 7.1 RPI-400, "Decontamination Program" 7.2 RPI-500, "Bioassay Program" 7.3 RPI-909, "Testing of Breathing Air System" 7.4 STA-625, "Foreign Material Exclusion" 7.5 STA-660, "Control of High Radiation Areas" 7.6 STA-682, "Control of Station Diving Operations" 7.7 STA-735, "Nuclear Fuel Integrity Program" 7.8 Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants 7.9 SER 10-97, "Unplanned Exposure During Spent Fuel Pool Diving Operations" 8.0 ATTACHMENTSIFORMS 8.1 Attachments Attachment 1, Pre-Dive Operations Check List Attachment 2, Instructions for Diving Operations Attachment 3, Post Dive Check List 8.2 EQoms None

69 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 8 OF 12 ATTACHMENT 1 PAGE 1 OF 1 PRE-DIVE OPERATIONS CHECK LIST UNIT [] 1 []2 Date SFP [I Transfer Canal [ I Reactor Cavity [ ] INITIAL IF YES 1) Has the underwater survey been completed? 2) The location of spent fuel and irradiated objects are verified on the survey and confirmed with Reactor Engineering. 3) Physical barriers have been erected, where appropriate, and are logged, in accordance with STA-735. I 4) Breathing air has been tested and meets the requirements of RPI-909, "Testing of Breathing Air Systems" or is certified by the vendor. 5) Underwater lights are operational. 6) The water is clear enough to maintain visual contact with the diver at the bottom of the pool. 7) Tritium analysis has been performed (divers and pool water). 8) Cameras checked. 9) Diver is aware of work area and diver path, and no deviation to established area will be granted. IF THEN 1-9 INITIALED YES ACTIVATE RWP ANY ANSWERED NO DO NOT ACTIVATE RWP

70 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 9 OF 12 ATTACHTENT 2 PAGE 1 OF 3 INSTRUCTIONS FOR DIVING OPERATIONS PARTA Date The following items shall be verified before the diver enters the water: 1) Has the underwater survey been completed? 2) The location of spent fuel and irradiated objects are verified on the survey and confirmed with Reactor Engineering. 3) Water clarity 4) Underwater lights are operational 5) Dosimetry placement and appropriately sealed 6) Dose rates in the dive area (brief diver using survey maps appropriately oriented to the dive area) 7) Diver allowable stay time (based on water temperature and dose rate) 8) Verify satisfactory results of daily Breathing Air Quality test.in accordance with RPI-909 9) Breathing air sufficient (have diver verify) 10) Communicatiors between the diver and tender (including agreement on a contingency plan for signaling the diver to leave the pool if communication is lost). 11) Cameras operational.

71 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 10 OF 12 ATTACHMENT 2 PAGE 2 OF 3 INSTRUCTIONS FOR DIVNG OPERATIONS PART B As the diver is being submerged, the following should be verified: 1) Communications between the diver and tender 2) Breathing air is sufficient (ask the diver) 3) Integrity of the suit (ask the diver) 4) Diver's well being 5) Dosimetry system operational If any of the above items are not suitable to the diver or RP Technician, the dive should be aborted. To ensure the safety of the diver while the dive is in progress, the RP Technician responsible for job coverage at the dive site should: 1) Maintain communications with the diver through the use of a head set or through the diver tender. 2) Maintain unobstructed visual contact with the diver. A viewing box is preferred, with no ripples, bubbles or other anomalies to prevent a clear view of the diver. Underwater cameras should also be used, where appropriate. 3) Monitor the dose rates at the work area and advise the diver of any significant changes.

72 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 11 OF 12 ATTACHMENT 2 PAGE 3 OF 3 INSTRUCTIONS FOR DIVIN OPERATIONS PART C (Continued) 4) Abort the dive if any of the following conditions are observed: a) interruption or loss of breathing air b) stay time has been reached c) loss of underwater lighting d) interruption of the ability to maintain clear visual contact e) leakage of water into the diver's dry suit f) loss of communications g) physical discomfort to the diver h) significant increase in radiation levels or unexpected dose rate encountered. i) failure of underwater radiation detectors j) failure of remote readout dosimetry system k) alarms (building evacuation or dosimetry) 1) diver leaves designated dive area (work area) m) camera failure J I

73 CPSES RADIATION PROTECTION INSTRUCTION MANUAL INSTRUCTION NO. RPI-611 RADIOLOGICAL CONTROLS FOR DIVING OPERATIONS REVISION NO. 3 PAGE 12 OF 12 ATTACHNMENT 3 PAGE I OF I POST DIVE CHECK LIST Date: At the end of the day or when diving operations have been suspended for an indeterminate period, the following should be addressed: 1) Return diver's dosimetry and obtain latest dose estimates 2) Store the diver's suit and equipment neatly 3) Close the valve to the breathing air system and store the hoses neatly 4) Close the valve to the demin water and store the hoses neatly 5) Contact the decon group for area decon and trash removal 6) FME controls

74 ENCLOSURE 2 to TXX Licensing Report for Spent Fuel Rack Installation at Comanche Peak Steam Electric Station Replacement Pages Pages 8-24 through , Table , Table , Table , Table 8.3-3

75 The axial force, bending moment and safety margin for the critical elements of the spent fuel pool under normal condition is given below: Axial Force* Bending Structural Kips Moment Safety Margin Component K-ft East Wall West Wall North Wall South Wall Slab *Tension Positive The axial force, bending moment and safety margin for critical elements of the spent fuel pool under factored load conditions is given below: Axial Force* Bending Structural Kips Moment Safety Margin Component K-ft East Wall West Wall North Wall South Wall Slab *Tension Positive It should be noted that the above results include a reduction in concrete strength due to high temperature and thermal cycling for the East and West walls. In addition, these results are based on the following conservative assumptions: " The maximum affects of gamma heating are applied to the concrete elements from the bottom of the pool, throughout the entire height of the racks, to a point several feet above the top of the racks. " The maximum wall temperatures, including the affects of gamma heating, are applied as indicated above around the entire perimeter of the pool. This would require that spent fuel from a full core off-load be placed in the outermost cell of every rack around the perimeter of the pool. In practice, newly off-loaded fuel is placed in the Region I racks. Therefore, temperatures down the length of the pool side walls and in the wall opposite the gate would be reduced. Page 8-24 Report HI

76 "* Thermal gradients through the pool walls assume that the back side of the wall is at minimum ambient temperature. "* Seismic accelerations for the top of the pool are used to compute seismic loads throughout the entire height of the pool. Evaluation of SFP2 Slab The bending moments, axial forces, and shear forces in the slabs and walls are evaluated for the governing load combinations from Section and then compared with capacity of the reinforced concrete cross sections. The evaluated safety margins (factors of safety), representing the ratio between the ultimate capacities of the cross-section and the computed internal force (moment), serve as direct indicators of structural integrity. In accordance with the ACI code (Reference 8.2.2), the structural evaluation uses the Ultimate Strength Design (USD) method for evaluation of the capacity of reinforced concrete members and considers the interaction of shear forces and bending moments with axial forces. The interaction between the axial force and bending moment is defined by the following parameters: Pure Compression Capacity (Po), Balance Point Capacity defined by axial force (Pb) and moment (Mb), Pure Bending Capacity (Mo), and Pure Tension Capacity (To). Each of these parameters represents a point in the Axial Force-Bending Moment (M-N) diagram. Two bending load cases are considered, referred to as positive moment capacity (when the moment causes tension in the outer reinforcement) and negative moment capacity (when the moment causes tension in the inner reinforcement), resulting in a total of six (6) reference points in the M-N diagram. Figure illustrates a typical M-N diagram. The ratio of the permissible moment to the actual moment computed by the finite element solution defines the factor of safety for bending. The permissible moment for specific location is obtained from the "M-N" diagram by using the in-plane force in an element computed by the finite element solution. For each element in the finite element model, two values of safety factors, corresponding to the two orthogonal planes of bending, are established for each factored load combination. Shear factors of safety are defined for each supported edge of the tunnel roof slab where the maximum shear stresses are generated. The distribution of the shear capacity along the edges of the slab is evaluated as a function of the in-plane force in accordance with Section 11.3 of ACI (Reference 8.2.2). The shear forces are then integrated along the total width of the cross section to obtain the total available shear capacity of the Page 8-25 Report HI

77 supporting edge. The in-plane forces used for shear capacity evaluation are obtained from the finite element solution. The total shear load is obtained by integrating the shear force resultants from the finite element solution throughout the total width of the support. The ratio of the total available capacity of supporting edge to the total shear load acting on the same reinforced concrete cross section defines the factor of safety for shear. The axial force, bending moment and safety margin for the critical element of the SFP2 slab are given below: Axial Force* Bending Structural Kips Moment Safety Margin Component K-ft Slab *Tension Positive The axial force, bending moment and safety margin for the critical element of the SFP2 slab under factored load conditions are given below: Axial Force* Bending Structural Kips Moment Safety Margin Component K-ft Slab *Tension Positive Local Structure Integrity Mechanical Accidents The maximum compressive stress calculated in the concrete as a result of the mechanical accident analysis is 11,374 psi. Per the ACI Code, the allowable bearing pressure for confined concrete (under static load) is 1.1 9fc. For concrete strength of 4000 psi, the allowable bearing pressure equals 4760 psi. Based solely on this static limit, one would infer that some concrete is crushed below the impact area. However, the deep drop event creates a scenario where the concrete is subjected to high strain rates. For this evaluation, application of the ACI Code to evaluate concrete limits is not appropriate. Under this loading condition, the concrete response can only be determined by application of an acceptable stress-strain relation for concrete. Since the concrete is confined laterally, the deep drop event causes a state of tri-axial compression in the concrete. A suitable model for concrete material subjected to tri-axial compressive loads is obtained from Appendix Page 8-26 Report HI Report HI

78 C of NUREG/CR-6608 (Reference 8.2.6). The model is defined by two yield vs. pressure curves, an upper (or undamaged) curve and a lower (or damaged) curve, so that the yield point can shift from the upper curve to the lower curve as the concrete fails in compression. When a finite element fails completely in LS-DYNA, the element is removed from the model by a process called material erosion. In the deep drop Scenario 2 analysis, all of the concrete elements remain present indicating that the stresses in the slab are below the failure limit. Therefore, structural integrity of the slab is maintained. Bearing Pads The average compressive stress in the concrete, calculated over the net effective interface area between the bearing pad and the concrete (determined while subject to the peak load) must remain below the allowable stress for confined concrete. The limit on bearing strength set by the ACI Code is used in the analysis. For concrete with a design compressive strength of 4000 psi, using ý = 0.7 and F = 1.0, the limit on bearing strength is 2380 psi. From the analysis, the maximum average compressive stress in the concrete is 1463 psi. Therefore, the structural integrity of the spent fuel pool concrete slab under maximum local bearing pressure from the rack pedestal is maintained. 8.3 Evaluation of SFP Stainless Steel Liner Description of SFP Liner The spent fuel pool liner consists of 3/16 inch thick stainless steel plate anchored into the concrete walls with headed Nelson studs with 12 inch center to center spacing each way. The liner plates are secured to the pool slab by welding to embedded strip plates which are, in turn, anchored with Nelson studs. The floor plates are a maximum of 40 inches wide and are continuously welded to the embedded plates. To protect these welds and maintain full contact area between the leveling pad assemblies and the pool floor, stainless steel shim bridges will be installed under the rack pedestals (bearing pads) when they are located over liner welds. Liner plate details are provided in Figures through The primary function of the liner is to provide a leak proof barrier for the spent fuel pool during fuel handling operations and spent fuel storage. Page 8-27 Report HI Report HI

79 8.3.2 Liner Analysis Gross Liner Structure Integrity The spent fuel pool liner must be evaluated for the additional loads imposed by the free standing racks. The original racks were anchored to the pool floor, so there was a direct load path from the rack to embedded anchor plates and then to the concrete structure. The high density racks are free standing, so lateral loads are resisted by skin friction in the liner. The strains due to skin friction must be combined with existing strains in the liner to determine total strain. The liner was previously evaluated for the effects of normal operating and accident temperatures which bound the current design temperatures. The liner analysis is based on a combination of hand calculations and computer analyses. Structural deformations and concrete reactions are obtained from the updated analysis of the spent fuel pool discussed in Section (one-half model of Fuel Building). Thus, the liner evaluation takes into consideration updated results from that evaluation Local Liner Structure Integrity Local structural integrity of the spent fuel pool liner is evaluated using results from the mechanical accident analyses described in Section Applicable Liner Codes Standards and Specifications Title 10 part 50, "Domestic Licensing of Protection Against Radiation" Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants" RG 1.13 Rev. 1 (12/75), "Spent Fuel Storage Facility Design Basis" Liner Load Definitions and Combinations Load Definitions The primary purpose of the spent fuel pool liners is to provide a water retaining barrier to maintain water levels during all normal and abnormal plant conditions. The liner acts as a membrane and is attached to supporting structure that is Seismic Category I and designed to structurally sustain normal and abnormal loading as outlined in Standard Review Plans through As such, the liner is not a structural member necessary for overall structural strength. However, the liner is designed to ensure its Page 8-28 Report HI Report HI

80 structural integrity during normal operation, earthquake, and other defined environmental and accident conditions as described herein. Water Pressure The liner and liner components are designed to withstand appropriate hydrostatic water pressures. The liner walls and floors are not specifically designed for water pressure, but are required to deform with the supporting concrete structure and remain watertight. Thermal Loads Consideration is given to the loads and stresses induced in the liner and liner components as a result of the pool water bulk thermal temperature and transient thermal effects. When determining thermal loads, it is assumed that the liner and concrete are initially an unstressed system at a temperature of 70 Degrees F. Normal operating conditions are evaluated and consideration must also be given to abnormal conditions, such as partial failure of the pool water cooling system. Thermal loading of the liner is the result of the difference in thermal expansion between the liner and the concrete structure to which it is attached. Neglecting deformations of the concrete structure due to other than thermal loads, the maximum load in the liner occurs when the temperature difference between the liner and concrete is maximum. Structural Deformation Attachment of the pool liner to the concrete structure results in strains being imposed on the liner caused by deformations of the concrete structure. These deformations are normally caused by dead loads, hydrostatic loads, live loads, seismic loads, and thermal gradients occurring in the concrete walls and floor slab. Consideration of the effects of the structural deformations and resultant liner strains in conjunction with other applicable loading effects are used in evaluation of the liner integrity. Normal Loads D = dead loads and their related moments and forces, including any permanent equipment loads. L = live loads and their related moments and forces, including any moveable equipment loads and other loads which vary with intensity and occurrence. Page 8-29 Report HI Report HI

81 TOM = thermal loads occurring as a result of normal operating conditions including the effects of gamma heating as applicable. Extreme Environmental Loads SSE = loads generated by the safe shutdown earthquake Abnormal/Extreme Environmental TOA = thermal loads occurring as a result of abnormal conditions, including TOM FD = loads generated from mechanical accident analysis Load Combinations The following load combinations are considered. Normal D + L + TOM Extreme Environmental D + L + ToM + SSE Abnormal/Extreme Environmental D + L + TOA + SSE D + L + TOM + FD Notes regarding load combinations: 1. Load factors are implicitly Severe environmental (OBE), abnormal, and abnormal/severe environmental load combinations are bounded by the above when load factors are D, L, and SSE loads acting on the wall liners are those resulting from the associated structural deformations and attached equipment reactions. 4. D, L, and SSE loads acting on the floor liner are those resulting from the spent fuel racks. 5. D, L, and SSE loads acting on embedments are those resulting from equipment reactions. 6. TOM and TOA loads are those resulting from the liner temperature increase above ambient at installation (i.e., 70 Degrees F). Page 8-30 Report HI

82 8.3.5 Description of Liner Analysis Gross Liner Structure Integrity Analysis of the spent fuel pool liner is based on a combination of hand calculations and computer analyses. Hand calculations are used to determine strain in the liner due to thermal expansion and friction loads from the rack pedestals. To determine the worst case location, the full array of pedestals is mapped on the pool floor. Loads induced in the liner due to structural deformations and concrete reactions are obtained from the updated spent fuel pool analysis described in Section (i.e., the one-half model of the Fuel Building) Local Liner Structure Integrity Local structural integrity of the spent fuel pool liner is evaluated using results from the mechanical accident analyses described in Section Liner Acceptance Criteria and Results Gross Liner Structure Integrity The design of spent fuel pool liners is not governed by a specific code other than the general requirements of NRC Regulatory Guide The design of CPSES SFP liners is based on the material strain limits from the ASME Boiler and Pressure Vessel Code, Section III, Division 2 (ACI Standard 359) for containment liners. This is very conservative since the pool liner material is stainless steel which has a much higher strain capacity than typical carbon steel containment liners. The extent of ASME criteria utilized is limited to the liner plate allowables presented in Table CC of the Code. Liner plate seam welds, other than fillets, are evaluated using the liner plate acceptance criteria. Liner plate allowables are presented in Table of this document. Liner plate fillet welds are evaluated in accordance with AISC rules with allowables increased by a factor of 1.5. Liner stud acceptance criteria is presented in Table Results of the liner evaluation are presented in Table Local Liner Structure Integrity The maximum strain in the liner resulting from a mechanical accident is in/in. This is less than the allowable tensile strain of in/in for this abnormal condition. Page 8-31 Report HI Report HI

83 8.4 Evaluation of Fuel Building Structure Design calculations for the Fuel Building structure were reviewed for impact due to the increased weight associated with high density spent fuel storage, as well as the new design floor response spectra. The existing building analyses were based largely on classical hand calculation analysis techniques. Because the pool floor location is at the base of the Fuel Building, the weight of the spent fuel storage racks and spent fuel assemblies has little impact on the overall building structure (i.e., loads are transferred to the pool slab, basemat, and founding soil). Thus, in the existing design basis analyses, the Fuel Building concrete beams, columns, slabs, and shear walls (excluding the spent fuel pool elements) were evaluated for loads from the overall building structure, including the weight of spent fuel pool concrete elements. These loads were distributed to/resisted by the reinforced concrete elements of the Fuel Building. In turn, the spent fuel pool structure was evaluated based on a finite element analysis that incorporated all of the Fuel Building structural elements through symmetry. To confirm that the Fuel Building structural elements were adequate for loads transmitted from the pool structure to the building elements, results from the finite element analysis were used to re-check Fuel Building structural elements. The load definitions and combinations in these analyses are described in Section As described in Section , to evaluate the affect of increased weight and hydrodynamic fluid structure interaction pressures, the spent fuel pool finite element analysis has been updated. Results of the updated finite element analysis have been used to re-check structural elements in the Fuel Building. These analyses demonstrate that structural elements in the balance of the Fuel Building remain within Code requirements. Seismic loads in the Fuel Building structural analyses are based on peak structure response (i.e., design floor response spectra ZPA). As documented in Section 8.1, the increased weight associated with high density spent fuel storage has no impact on peak structure responses. 8.5 Evaluation of Fuel Building Internal Structures, Systems, and Components The affect of the revised building responses on the balance of Fuel Building structures, systems, and components has been evaluated. These evaluations demonstrate that the design of the Fuel Building structures, systems, and components remains adequate considering the changes in design response spectra and that no modifications are necessary Piping Systems Piping systems and components inside the Fuel Building are classified as being either safety related or non-safety related. The safety related piping systems are ASME Class 2 and 3, and are further classified as being Seismic Category I. The Page 8-32 Report HI

84 non-safety related piping systems are classified as Seismic Category II. Regardless of the safety classification, reference to the piping system includes the piping, supports and all in-line components Seismic Category I Piping Systems As stated in FSAR Section 3.7B. 1.3, all safety related and Seismic Category I piping in the Fuel Building has been designed and analyzed using Code Case N-411 damped spectra. This applies to both large bore piping and small bore piping. The only spectral exceedances identified in the new N 411 damped design floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur at Elevations ft and 918 ft. A complete review of piping systems located in the Fuel Building has been performed and it was concluded that there is no safety related piping at these elevations. Therefore, the changes in floor response spectra have no impact on safety related piping Seismic Category II Piping Systems Large and small bore Seismic Category II piping at CPSES installed prior to receiving the Unit 1 Operating License was evaluated based on a combination of walkdowns, analyses, and assessments using earthquake experience and test data. The earthquake experience based approach focused on critical piping attributes that lead to piping failures in strong motion earthquakes. Plant walkdowns were used to identify piping systems with critical attributes and to select bounding or representative piping system configurations for more detailed analysis. For small bore piping, twenty sampling calculations were performed on critical piping systems located throughout the plant. Three of the sampling calculations pertain to piping located in the Fuel Building. The sampling analyses were performed using N-411 damped spectra. As stated in Section , the only spectral exceedances identified in the new N-41 I damped design floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur at Elevations ft and 918 ft. One of the three small bore piping sampling calculations for the Fuel Building pertains to piping located below Elevation 860 ft. Therefore, this sampling calculation is not impacted by the change in design response spectra. The remaining two small bore piping sampling calculations have been reviewed. The response spectra used in these evaluations is an envelope of Fuel Building and Auxiliary Building spectra at the upper elevations of the buildings. The enveloped spectra was reviewed and it was determined that Page 8-33 Report HI Report HI

85 there is no impact due to the changes in the Fuel Building design response spectra at Elevations ft and 918 ft. For large bore piping, ten sampling calculations were performed on critical piping systems located throughout the plant. Only one of the sampling calculations pertains to piping located in the Fuel Building. The sampling analyses were performed using N-41 1 damped spectra. The large bore piping sampling calculation for the Fuel Building pertains to piping located below Elevation 860 ft. Therefore, this sampling calculation is not impacted by the change in design response spectra. Most Seismic Category II small bore piping installed after receipt of the Unit 1 Operating License is based on standard (cookbook) details. Seismic loads for the standard details are based on peak accelerations enveloped for multiple building locations and elevations. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, piping systems installed in accordance with the standard (cookbook) details are not impacted by changes in the design floor response spectra. All Seismic Category II large and small bore piping systems installed in the Fuel Building after receipt of the Unit 1 Operating License were reviewed. No new/modified piping designed by the response spectrum method has been designed/installed in the upper elevations of the Fuel Building (i.e., areas where re-rack spectra exceeds design spectra). Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category II piping systems Floor Mounted Components Floor mounted components include both plant and non-plant equipment. Non-plant equipment includes components that are not directly required to operate the plant Seismic Category I Floor Mounted Components Seismic Category I floor mounted components at CPSES are qualified by analysis, testing, or a combination of analysis and testing. Floor mounted components are those components (or equipment) that are mounted to floors, walls, or the under side of slabs, including their supports. This category of components does not include in-line components, such as valves. In-line components are considered as part of the supporting system. Active components are qualified using 2% damped OBE response spectra and 2% damped SSE response spectra. Non-active components are qualified using 2% damped OBE response spectra and 3% damped SSE response spectra. Welded steel equipment supports are qualified using 2% Page 8-34 Report HI

86 damped OBE response spectra and 2% damped SSE response spectra for active components and 4% damped SSE response spectra for non-active components. Bolted steel equipment supports are qualified using 4% damped OBE response spectra and 4% damped SSE response spectra for active components and 7% damped SSE response spectra for non-active components. The only spectral exceedances identified in the new OBE 2% and 4% and SSE 2%, 3%, and 4% damped design floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur in the horizontal (E-W and N-S) directions at Elevations 860 ft, ft, and 918 ft. For 7% damped SSE response spectra, exceedances occur in the horizontal (E-W and N-S) directions at Elevations ft and 918 ft. A complete review of Seismic Category I floor mounted components has been performed to determine the impact of the changes in response spectra at these locations. Due to the layout and configuration of the Fuel Building (e.g., spent fuel pool operating deck is located at elevation 860 ft), there are minimal Seismic Category I components at/above Elevation 860 ft. The most significant components at the affected elevations of the Fuel Building are the overhead crane and spent fuel pool level instrumentation. Seismic qualification documentation for these components has been reviewed and it has been determined that these components remain qualified considering the new design response spectra. Components located below Elevation 860 ft which use the 860 ft elevation response spectra for seismic qualification were also reviewed. The most significant components in these areas are the new fuel storage racks, spent fuel pool / transfer canal gates, and inclined transfer tubes. Seismic qualification documentation for these components has been reviewed and it has been determined that these components remain qualified considering the new design response spectra. In addition to permanent plant equipment, the qualification of the temporary crane (a.k.a. "Wonderhoist") used during installation of the high density spent fuel storage racks was reviewed. Qualification of the Wonderhoist is based on 7% damped SSE response spectra at Elevation 860 ft, since the SSE condition envelopes the OBE condition. There are no changes in the 7% damped SSE response spectra at Elevation 860 ft. Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category I floor mounted components Seismic Category II Floor Mounted Components Seismic Category II floor mounted components at CPSES installed prior to receiving the Unit 2 Operating License were evaluated based on a Page 8-35 Report HI-2(J(J'2402 Report HI

87 combination of walkdowns, analyses, and assessments using earthquake experience and test data. The earthquake experience based approach focused on critical attributes that lead to component anchorage/support failures in strong motion earthquakes. Plant walkdowns were used to identify components with critical attributes and to select bounding or representative configurations for more detailed analysis. The earthquake experience based approach identified twenty-four (24) components in the Fuel Building as candidates for further evaluation. Of these twenty-four candidates, ten (10) are located in exclusion areas. Exclusion areas have no, or limited, safety related commodities. Therefore, non-seismic components in exclusion areas are maintained under the Seismic Interaction Program which monitors the installation of new safety related equipment. No new safety related equipment has been installed in proximity of these ten components. Since these components are not designed for seismic loads, the changes in Fuel Building dynamic response discussed in Section 8.1 have no impact on their qualification. Eleven (11) of the remaining items are located at or below Elevation 810'-6". There are no changes to the response spectra used to qualify these items. The remaining three (3) items were reviewed and it was determined that they are not impacted by the changes in design floor response spectra. Therefore, all Seismic Category II components remain qualified without modification. All Seismic Category II floor mounted components installed in the Fuel Building after receipt of the Unit 1 Operating License were reviewed. No new/modified components have been installed that used seismic accelerations affected by the revised design floor response spectra for the upper elevations of the Fuel Building. As noted in the discussion in Section , the only spectral exceedances identified in the revised floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur in the horizontal (E-W and N-S) directions at Elevations 860 ft and above. One significant Seismic Category II component located at Elevation 860 ft, the fuel handling bridge crane, was evaluated reviewed in additional detail. Based on a review of existing seismic qualification documentation for the bridge crane, it was determined that the crane qualification is not impacted by the new design response spectra. Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category II floor mounted components. Page 8-36 Report HI Report HI

88 8.5.3 Electrical Raceways Electrical raceways consist of power, I&C and control cables that are routed in either conduit or cable tray. Also included with the raceways are their associated supports Seismic Category I Electrical Raceways Conduit Systems Three basic methods are used to determine seismic loads on Seismic Category I conduit raceways; static analysis, equivalent static analysis, and response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on spectra with 2% damping for the Operating Basis Earthquake (OBE) and 3% damping for the Safe Shutdown Earthquake (SSE). For conduit covered with thermolag fire protection material, spectra with 4% damping for the OBE and 7% damping for the SSE can be used. Seismic loads for static analyses are based on peak accelerations with a multi-mode factor of 1.5. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, conduit raceways evaluated using static analysis are not impacted by changes in the design floor response spectra. Seismic loads for equivalent static analyses are based on peak spectral accelerations for system frequencies at or to the left of frequencies corresponding to the peak (i.e., the low frequency side). For frequencies to the right of the peak (i.e., the high frequency side), the acceleration at the corresponding frequency is applied. This is done for each direction of input. In all cases, an appropriate multi-mode factor is applied when the system frequency is less than 33 Hz. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on conduit raceways evaluated using the equivalent static analysis approach since peak accelerations have been used to evaluate conduit systems with frequencies at or below the peak. Seismic loads for response spectrum analyses are based on modified design floor response spectra. The spectra was modified such that the response spectra accelerations on the left side of the peak were set equal to the peak accelerations for both OBE and SSE. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on conduit systems evaluated using the Page 8-37 Report HI J2

89 response spectrum method of analysis since peak accelerations have been used for systems with frequencies at or below the peak. Cable Tray Systems Three basic methods are used to determine seismic loads on Seismic Category I cable tray raceways; static analysis, equivalent static analysis, and response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on spectra with 4% damping for the Operating Basis Earthquake (OBE) and 7% damping for the Safe Shutdown Earthquake (SSE). Seismic loads for static analyses are based on peak accelerations with a multi-mode factor of 1.5. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, cable tray raceways evaluated using static analysis are not impacted by changes in the design floor response spectra. Seismic loads for equivalent static analyses are based on peak spectral accelerations for system frequencies at or to the left of frequencies corresponding to the peak. For frequencies to the right of the peak, the acceleration at the corresponding system frequency is applied. This is done for each direction of input. In all cases, an appropriate multi-mode factor is applied for systems with a fundamental frequency below 33 Hz. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on cable tray raceways evaluated using the equivalent static analysis approach since peak accelerations have been used to evaluate cable tray systems with frequencies at or below the peak. Seismic loads for cable tray systems qualified by response spectrum analyses are based on the design floor response spectra. As stated above, only the spectra at elevations ft and 918 ft are affected by the revised building response. All cable tray system analyses that used the response spectrum method of analysis have been reviewed. None of the cable tray systems that are evaluated using the response spectrum method are located above elevation 860 ft. Therefore, there is no impact Seismic Category II Electrical Raceways Conduit Systems 2" and Under (Small Bore) Seismic Category II small bore conduit systems installed prior to receipt of the Unit 1 Operating License were evaluated using a multi-level screening process based on the requirements set forth in Standard Review Plan Page 8-38 Report HI

90 The approach verifies that Seismic Category II small bore conduit supports maintain structural integrity, do not interact with safety-related plant features, or do not have adverse interaction with safety-related plant features if the conduit system fails. The revised building responses discussed in Section 8.1 have no affect on conduits evaluated by interaction. Most of the Seismic Category II conduit systems evaluated for seismic structural integrity consider peak accelerations. However, plant procedures allow response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on spectra with 7% damping for the Safe Shutdown Earthquake (SSE). The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at Elevations ft and 918 ft. These exceedances are localized and would not significantly alter seismic loads. At worst, previously calculated responses would be underestimated in a single mode under consideration. Based on plant walkdowns, the conduit systems located above Elevation 860 ft are supported with standard (cookbook) supports found to be seismically robust in previous evaluations. Therefore, conduit system qualification is not impacted by the changes in design floor response spectra. Seismic Category II small bore conduit systems installed after receipt of the Unit 1 Operating License are in accordance with standard (cookbook) details. Seismic loads for the standard details are based on peak accelerations (with an appropriate multi-mode factor) enveloped for multiple building locations and elevations. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, conduit systems installed in accordance with the standard (cookbook) details are not impacted by changes in the design floor response spectra. Conduit Systems Greater than 2" (Large Bore) Most of the Seismic Category II conduit systems greater than 2" diameter are evaluated using the same methods as Seismic Category I conduit systems. Accordingly, the discussion and conclusions for Seismic Category I conduit systems apply to the greater than 2" diameter Seismic Category II conduit systems. A fraction of the Seismic Category II conduit systems greater than 2" diameter are not seismically supported. These conduits are either restrained from falling during a seismic event by means of aircraft cable restraints or are supported solely by dead weight supports. The design of aircraft cable restraints is based on peak seismic accelerations. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. The conduits supported solely by dead weight supports were evaluated by the System Interaction Program for possible interaction with Seismic Page 8-39 Report HI

91 Category I systems, structures, and components (SSC's). Thus, these systems are not impacted by changes in the design floor response spectra. Therefore, the revised building responses discussed in Section 8.1 have no affect on the qualification of Seismic Category II conduit systems greater than 2" diameter. Cable Tray Systems Seismic Category II cable tray systems are evaluated using the same methods as Seismic Category I cable tray systems. However, only consideration of the Safe Shutdown Earthquake (SSE) is required. The discussion and conclusions for Seismic Category I cable tray systems apply to the Seismic Category II cable tray systems I&C Tubing Instrumentation and Control tubing systems inside the Fuel Building are classified as either safety related or non-safety related. The safety related tubing systems are ANSI Safety Class 2 and 3, and are further classified as being Seismic Category I. The non-safety related tubing systems are classified as Seismic Category II. Regardless of the safety classification, reference to the tubing system includes the tubing, supports and all in-line components. Also included in the scope of I&C Tubing are wall and floor mounted instruments and instrument supports Seismic Category I I&C Tubing Tubing and Supports Seismic Category I I&C tubing systems in the Fuel Building have been designed and analyzed using Code Case N-411 damped spectra. The only spectral exceedances identified in the new N-41 1 damped design floor response spectra (i.e., envelope of existing design response spectra and re rack spectra) occur at Elevations ft and 918 ft. A complete review of tubing systems located in the Fuel Building has been performed and there is no safety related tubing at these elevations. Therefore, the changes in floor response spectra have no impact on safety related tubing systems. Instrument Supports Seismic Category I instruments at CPSES are typically supported using standard (typical) pre-qualified support details. The typical supports are designed to maintain natural frequencies in the rigid range of the response spectra. Instrument supports are designed and analyzed using 2% OBE and 3% SSE response spectra. The only spectral exceedances identified in the Page 8-40 Report HI

92 new 2% OBE and 3% SSE damped design floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur in the horizontal (E-W and N-S) directions at Elevations 860 ft, ft, and 918 ft. None of these exceedances affect the rigid range spectral accelerations. Therefore, there is no impact to instruments supported using the standard pre-qualified support details. All of the Seismic Category I instruments were further reviewed to identify instruments with unique support designs. None of the instruments in the affected areas of the Fuel Building are installed using unique support designs. The changes in design response spectra have no impact on the seismic qualification of Seismic Category I instrument supports Seismic Category II I&C Tubing Tubing and Supports All Seismic Category II I&C tubing systems are field routed and installed in accordance with site approved procedures and standard (cookbook) span limitations and support details. Seismic loads for the cookbook details are based on peak accelerations enveloped for multiple building locations and elevations. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, Seismic Category II I&C tubing and supports are not impacted by changes in the design floor response spectra. Instrument Supports Seismic Category II instruments at CPSES are typically supported using standard (typical) pre-qualified support details. The typical supports are designed and analyzed using 7% SSE response spectra. The only spectral exceedances identified in the new 7% SSE damped design floor response spectra (i.e., envelope of existing design response spectra and re-rack spectra) occur in the horizontal (E-W and N-S) directions at Elevations ft and 918 ft. None of these exceedances affect peak spectral accelerations. Therefore, there is no impact to instruments supported using the standard pre-qualified support details. All of the Seismic Category II instruments were further reviewed to identify instruments with unique support designs. None of the instruments in the affected areas of the Fuel Building are installed using unique support designs. Page 8-41 Report HJ Report HI

93 Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category II instrument supports HVAC Systems HVAC Systems consist of air handling duct, duct supports, and in-line hardware such as turning vanes, registers, and dampers. Also included within the scope of HVAC Systems are air handling units, plenums, and HVAC equipment supports Seismic Category I HVAC Systems Duct and Supports There are no Seismic Category I HVAC duct systems supported from the Fuel Building structure. Air Handling Units, Plenums, and HVAC Equipment Supports (AHPES) Three methods were used to determine seismic loads on Seismic Category I air handling units, plenums, and equipment supports; static analysis, equivalent static analysis, and response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on spectra with 4% damping for the Operating Basis Earthquake (OBE) and 7% damping for the Safe Shutdown Earthquake (SSE). Conservatively, 2% OBE spectra and 3% or 4% SSE spectra could be used. Seismic loads for static analyses are based on peak accelerations with an appropriate multi-mode factor. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, air handling units, plenums, and equipment supports evaluated using static analysis are not impacted by changes in the design floor response spectra. Seismic loads for equivalent static analyses are based on peak spectral accelerations when the fundamental frequency is at or to the left of frequencies corresponding to the peak. For frequencies to the right of the peak, the acceleration at the corresponding frequency is applied. This is done for each direction of input. In all cases, an appropriate multi-mode factor is applied. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on air handling units, plenums, and equipment supports evaluated using the equivalent static analysis approach since peak accelerations have been used to evaluate conduit systems with frequencies at or below the peak. Page 8-42 Report HI Report HI

94 Seismic loads for response spectrum analyses are based on the design floor response spectra profile. All air handling units, plenums, and equipment supports that were evaluated using the response spectrum method of analysis have been reviewed. None of the air handling units, plenums, and equipment supports evaluated using the response spectrum method of analysis are located in the upper elevations of the Fuel Building. Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category I HVAC systems Seismic Category II HVAC Systems Duct and Supports Seismic analysis of Seismic Category II duct systems is performed using either static analyses or equivalent static analyses. In accordance with the plant design basis, only the Safe Shutdown Earthquake (SSE) is required to be considered. The design accelerations are based on spectra with 7% damping (SSE spectra with 3% damping can conservatively be used). Seismic loads for static analyses are based on peak accelerations with a multi-mode factor of 1.5. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, HVAC duct systems evaluated using static analyses are not impacted by changes in the design floor response spectra. Seismic loads for equivalent static analyses are based on peak spectral accelerations for system frequencies at or to the left of frequencies corresponding to the peak. For frequencies to the right of the peak, the acceleration at the corresponding frequency is applied unless the fundamental frequency is above 33 Hz. This is done for each direction of input. In all cases, an appropriate multi-mode factor is applied. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on duct systems evaluated using the equivalent static analysis approach since peak accelerations have been used to evaluate systems with frequencies at or below the peak. No other exceedances were noted in the 7% damped SSE response spectra curves. Therefore, the revised building responses discussed in Section 8.1 have no affect on Seismic Category II HVAC systems evaluated by the equivalent static analysis method. Air Handling Units, Plenums, and HVAC Equipment Supports (AHPES) Three methods were used to determine seismic loads on Seismic Category II air handling units, plenums, and equipment supports; static analysis, Page 8-43 Report HI

95 equivalent static analysis, and response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on 7% damping for the Safe Shutdown Earthquake (SSE). Conservatively, 3% or 4% SSE spectra could be used. The discussion and conclusions for Seismic Category I air handling units, plenums, and equipment supports applies to the Seismic Category II AHPES Internal Structures Internal structures consist of access platforms, stairways and ladders, and grating Seismic Category I Internal Structures There are no Seismic Category I internal structures in the Fuel Building Seismic Category II Internal Structures Three methods were used to determine seismic loads on Seismic Category II internal structures; static analysis, equivalent static analysis, and response spectrum analysis. In accordance with the plant design basis, seismic accelerations are based on spectra with 4% damping for the Operating Basis Earthquake (OBE) and 7% damping for the Safe Shutdown Earthquake (SSE). Seismic loads for static analyses are based on peak accelerations with an appropriate multi-mode factor. The revised building responses discussed in Section 8.1 have no affect on peak accelerations. Therefore, internal structures evaluated using static analysis are not impacted by changes in the design floor response spectra. Seismic loads for equivalent static analyses are based on peak spectral accelerations when the fundamental frequency is at or to the left of frequencies corresponding to the peak. For frequencies to the right of the peak, the acceleration at the corresponding frequency is applied. This is done for each direction of input. In all cases, an appropriate multi-mode factor is applied. The revised building responses discussed in Section 8.1 do not affect peak accelerations, however, the spectra shift to the left in both horizontal directions at elevation ft and 918 ft. These shifts have no impact on internal structures evaluated using the equivalent static analysis approach since peak accelerations have been used to evaluate structures with frequencies at or below the peak. Seismic loads for response spectrum analyses are based on the design floor response spectra profile. All internal structures that were evaluated using Page 8-44 Report HI

96 8.6 Conclusions the response spectrum method of analysis have been reviewed. None of the internal structures evaluated using the response spectrum method of analysis are located in the upper elevations of the Fuel Building. Therefore, the changes in design response spectra have no impact on the seismic qualification of Seismic Category II internal structures. The CPSES Fuel Building and associated systems, structures, and components have been evaluated for the addition of high density spent fuel storage racks. Results of the investigations described in this report demonstrate that the building response is not sensitive to the added mass. One noticeable difference between the existing FSAR based floor response spectra and the new re-rack spectra is that peak accelerations in the re-rack spectra are lower in all cases. A second difference is that resonant frequencies (amplified building response) have shifted slightly to the left in the horizontal directions at the upper building elevations (899.5 ft and 918 ft). This affect is not attributable to installation of the high density spent fuel storage racks. Rather, it is the result of refinements in the mathematical model. The zero period acceleration of the re-rack floor response spectra are consistently below existing FSAR based floor spectra values. The spent fuel pool structures have been evaluated for the addition of high density spent fuel storage racks. Both gross structural integrity and local structural integrity were investigated. Gross structural integrity of the spent fuel pools was previously evaluated for installation of high density spent fuel storage racks in support of License Amendment Request These analyses were updated to incorporate revised interface loads from Holtec International (manufacturer for Region I racks in both pools and modifier of the Region II racks in SFP 1). Results of this analysis demonstrate that all of the structural elements of the spent fuel pools remain qualified considering the additional loading (including the effects of gamma heating) from high density spent fuel storage. To evaluate the reduced thickness of the slab in SFP2, a second analysis was prepared by Holtec. Results of this analysis demonstrate that structural elements of the SFP2 slab remain qualified considering the additional loading (including the effects of gamma heating) from high density spent fuel storage. Local structural integrity of the spent fuel pool slab has been assured by evaluating the consequences of fuel assembly drop accidents. In addition to the drop accidents, the pool slab has been evaluated for the maximum postulated loads from the rack pedestals (pads) for all loading conditions that do not include postulated mechanical accidents. In both cases, analyses demonstrate that the pool slabs will maintain structural integrity in the high density spent fuel storage configuration. Page 8-45 Report HI Report HI

97 In addition to the spent fuel pool reinforced concrete elements, the spent fuel pool liner was evaluated for the additional loads imposed by the free standing racks. Again, both gross liner integrity and local liner integrity were evaluated. Lateral loads from the free standing racks are resisted by skin friction in the liner. The resulting strains have been combined with strains from normal and accident load conditions to determine total strain in the liner. Results of the analysis demonstrate that total strains remain within acceptable limits. Local structural integrity of the spent fuel pool liner was evaluated using results from the mechanical accident analyses. Results of this evaluation demonstrate that the liner will maintain its ability to provide a leak tight barrier. Results of the updated Fuel Building structural analyses demonstrate that there is no adverse impact on loading in the balance of the Fuel Building structural elements due to the additional weight and hydrodynamic fluid structure interaction pressures. Where necessary, existing calculations for the Fuel Building have been revised to incorporate the new loads. All concrete elements remain within Code requirements. The affect of the revised building responses on the balance of Fuel Building structures, systems, and components has been evaluated. Results of this evaluation show that the design of the Fuel Building structures, systems, and components remains adequate and that no modifications are necessary. Page 8-46 Report HI-20U24U2 Report HI

98 Table AXIAL FORCE AND MOMENT FOR SLAB ELEMENTS Structural Elem Load *Axial Force Bending Moment Components # Direction (kips/ft) (k-ft/ft) Normal Load Condition Mat 471 -Y (E-W) Mat 472 -Y (E-W) Mat 473 -Y (E-W) Mat 474 -Y (E-W) Mat 463 -Y (E-W) Accident Load Condition Mat 471 -Y (E-W) Mat 472 -Y (E-W) Mat 473 -Y (E-W) Mat 474 -Y (E-W) Mat 463 -Y (E-W) * Tension positive. Page 8-58 Report HI Report HI

99 Table AXIAL FORCE AND MOMENT FOR WALL ELEMENTS Structural Elem Load *Axial Force Bending Moment Components # Direction (kips/ft) (k-ft/ft) Normal Load Condition East Pool Wall 24 -X (hor.) North Pool Wall 141 -Y (vert.) West Pool Wall 244 +X (hor.) West Pool Wall 245 +X (hor.) South Pool Wall 301 -X (hor.) South Pool Wall 302 -X (hor.) South Pool Wall 303 -X (hor.) South Pool Wall 341 -Y (vert.) Accident Load Condition East Pool Wall 47 -X (hor.) North Pool Wall 105 -X (hor.) West Pool Wall 244 -X (hor.) West Pool Wall 243 -X (hor.) South Pool Wall 301 -X (hor.) South Pool Wall 302 -X (hor.) South Pool Wall 303 -X (hor.) South Pool Wall 341 -Y (vert.) * Tension positive. Page 8-59 Keport 1-ll-ZUU24UL Report HI-20024UZ

100 TABLE SPENT FUEL POOL LINER ANCHOR ALLOWABLES Category Force/Displacement Allowable Mechanical Loads Displacement Limited Loads Lesser of: Service Loads Fa = 0.67Fy 6a = u Fa = 0.33Fu Factored Loads Lesser of: Fa = 0.90Fy 8a = u Fa = 0.50Fu I Where: Fa Fy Fu 8a 8u is the allowable load on the anchor stud without consideration of stud spacing. Fa is reduced for closely spaced studs. is the anchor stud yield capacity based on vendor data. is the anchor stud ultimate capacity based on vendor data. is the allowable displacement. is the ultimate displacement based on vendor data. Page 8-61 Report HI Report HI

101 TABLE SUMMARY OF MAXIMUM STRAINS/DISPLACEMENTS IN CRITICAL COMPONENTS OF THE SPENT FUEL POOL LINER/ANCHORAGE Liner Welds Studs Load Combination Actual Strain Actual Strain Actual Strain (Allowable Strain) (Allowable Strain) (Allowable Strain) Normal in/in in in (D+L+ToM) ( in/in) ( in) ( in) Extreme Environmental in/in in in (D+L+ToM+SSE) ( in/in) ( in) ( in) Abnormal Thermal/Extreme Environmental in/in in in (D+L+ToA+SSE) ( in/in) ( in) ( in) Page 8-62 Report HI Report HI

102 ENCLOSURE 3 to TXX Westinghouse Letter logged WPT-16160, Rev.1 from J. S. Wyble to C. L. Terry dated December 6, 2000 "NSAL , Axial Burnup Shape Reactivity Bias

103 0 Westinghouse Box 355 Electric Company Pittsburgh Pennsylvania December 6, 2000 WPT-16160, Rev. 1 Mr. C. L. Terry, Senior Vice President & Principal Nuclear Officer TXU Electric Company P.O. Box 1002 Glen Rose, TX Attention: VETIP Coordinator (No Response Required) TXU Electric Company Comanche Peak Steam Electric Station Unit Number 1 & 2 NSAL Axial Burnup Shape Reactivity Bias Dear Mr. Terry: The subject Nuclear Safety Advisory Letter is forwarded for your information and use. This letter provides the Westinghouse conclusion regarding 10 CFR 21 reportability, plant applicability, safety significance, and recommended actions. Westinghouse has determined that this issue is not reportable pursuant to the requirements of 10 CFR Part 21. The subject Nuclear Safety Advisory Letter addresses a non-conservatism in the axial burnup bias calculation which is part of the methodology used by Westinghouse for calculating criticality of spent fuel pool configurations. Westinghouse has determined that we provided the analysis of spent fuel pool criticality in your plant. Attachment 2 is the result of a plant specific evaluation of the magnitude of the credits for conservatisms that may be considered to compensate for the non-conservatism. Attachment 2 and the information in the NSAL are provided so that you may evaluate what actions or review are required to assure regulatory compliance. At the request of TXU, additional information to supplement the attached NSAL has been requested. In accordance with this request, the following clarifications and supplemental comments are provided. The attached NSAL contains background on the evolution of the Axial Burnup Bias issue leading to this notification. In addition, the NSAL provides a description of the methods used to evaluate the issue. Potential non-conservatisms were identified by Duke Power based on a preliminary study with an alternate methodology, and were brought to the attention of the NRC in March In response to this information, Westinghouse deemed it prudent to open a Potential Reportable Technology Error (PRTE),

104 WPT-16160, Rev. 1 2 December 6, 2000 also in March, to investigate potential applicability to the Westinghouse criticality methodology of WCAP NP-A, Revision 1. In May 2000 sufficient information had been gathered to confirm that concerns related to the conservatism of the Axial Burnup Bias as applied in WCAP NP-A were substantive. Consistent with Westinghouse/Nuclear Fuel internal procedures, a meeting of the Conditions Adverse To Safety (CATS) committee was requested to evaluate the appropriate reclassification of the issue. Recommendations were offered that the significance of the issue warranted elevation of the PRTE to a Potential Safety Issue (PI) pursuant to 10 CFR 21 and consistent with the Westinghouse Quality Management System. The CATS committee and the Westinghouse Safety Review Committee (SRC) -concurred; and-pk-00 t12 was opened-shortlyhaereafter-, in-may consistenrwith-wog-agreme-nits, the issue was summarized for review with the WOG Potential Issue (PI) Core Team. Based on their review, the team requested review prior to release of the formal notification of resolution of the P1. With sufficient basis established from the evaluation process, the Part 21 aspect of PI was closed within the 60-day "discovery" period allowed by Part 21 with a conclusion that no significant safety hazard existed as a result of the identified non-conservatisms in the methodology. A Nuclear Safety Advisory Letter (NSAL) was developed to provide notification of the issue, conclusions of the evaluation, and other pertinent information to the potentially affected plants. Upon review and reconciliation of comments with the WOG PI Core Team, distribution of the NSAL was processed via Project Letter in accordance with standard Westinghouse practice. It should be noted that PI was left open pending the issuance of the NSAL. This process facilitated tracking of the issue and issuance of the NSAL. The potentially affected analyses performed by Westinghouse for Comanche Peak are: A. "Spent Fuel Storage Rack Final Criticality Report," WPT , Nov. 29, [Note: Analysis for the 1/4 and 2/4 checkerboard configurations. No Boraflex panels. Soluble boron not credited. Outer wrapper not credited.] B. "Comanche Peak Region 2 Spent Fuel Rack Criticality Analysis Using Soluble Boron Credit, Final Report," 98TB-G-0016, Nov. 10, 1998 (Report CAC ). [Note: Analysis for the 3/4 and-4/4 storage configurations. No Boraflex panels. Soluble boron credited. Outer wrapper credited.] C. "Spent Fuel Rack Criticality Analysis," CAB per WPT-16105, May 5, [Note: Re-analysis of(b) except the outer wrapper is not credited.] The NSAL concludes that all potentially affected plants, including Comanche Peak, are technically acceptable for their current Technical Specifications spent fuel pool configurations less questions surrounding decay time credit as discussed below. The NSAL also includes a summary of the evaluations performed to reach this conclusion. The Westinghouse position of these evaluations is that they may serve as guidance for the utility in their assessment of continued applicability of this conclusion for future activities. Note that no attempt was made to customize the evaluations, merely to demonstrate acceptability. Variance from these evaluations, therefore, does not necessarily imply the need for immediate identification of new limits, spent fuel pool fuel shuffling, and/or reanalysis since additional margins may exist for any given plant or configuration change. In using the approach described in the NSAL, no new limits are established. Although

105 WPT-16160, Rev. 1 3 December 6, 2000 quantification of margins is not readily identifiable from this approach, the conclusions support that continued conformance to current limits (contingent on the evaluations) supports the conclusion of technical acceptability. It is also important to note that this is a "methodology" issue and has no safety implications in light of the conclusion of technical acceptability. Westinghouse makes no recommendations with regards to licensing actions as it relates to the approach used for-resolution of this issue-. Licensing activities beyond the conclusion-of-technical--aceptability-are left to the discretion of the license holder. It is recognized that TXU is currently holding one analysis that has been approved, but not implemented, and one analysis that has not yet been reviewed. On October 19, 2000, a conference call was held between the US-NRC, Westinghouse and utility representatives. The purpose of the conference call was to officially inform the NRC of Westinghouse's findings regarding the Axial Burnup Bias; appropriate credits to offset the bias penalty to demonstrate that the issue was of low safety significance; and to discuss the compliance issue related to the topical report. Applicability of Generic Letter 91-18, Revision 1 was also discussed. The NRC did not verbalize any concerns with the applicability of GL 91-18, Revision 1 and agreed that this approach was appropriate for all reviewed and approved analyses. For analyses under review, it was noted that the methodology issue needs to be explicitly addressed. All known issues related to the Westinghouse criticality methodology of WCAP NP-A, Revision 1 are addressed via this NSAL. It should be noted that a question has been raised with regards to Decay Time Credit and the impact of the axial burnup bias. This question is being reviewed within Westinghouse and any potential impacts from this review will be handled in accordance with the Westinghouse QMS and commensurate with the safety significance. If you have any questions concerning this Advisory Letter, please contact William Slagle at either or slaglewh@westinghouse.com, or Donald Lindgren at or lindgl da@westinghouse.com. Very truly yours, Attachment 1 - NSAL Attachment 2, Rev. 1 J. S. WybleC f. Customer Projects Manager

106 WPT-16160, Rev. 1 4 December 6, 2000 cc: C.L. Terry -1L, 1A M. R. Blevins - 1L, 1A S. Smith- 1L, 1A R. Flores- IL, 1A D. J. Reimer - 1L, IA VETIP Coordinator - IL, 1A CCG, 006A- 1L, IA M. R. Kilgore - 1L, 1A - IL, IA J. Kelley- 1L, 1A

107 WPT-16160, Rev. 1 5 December 6, 2000 bcc: J. S. Wyble/File Copy (WM F Bldg, M.B. 30) - 1L, 1A S. J. Hyde (WM F Bldg, M.B. 30) - 1L, 1A H. W. Gutzman (TBX Site) - IL, 1A S. Riggs (Dallas Sales) - IL, 1A M. A. Olson (WM F Bldg, M.B. 30) - 1L, la J. L. Vota (WEC E5-10) - 1L, 1A V. Polizzi (TBX Site) - 1L, IA

108 ATTACHMENT 1 Westinghouse Electric Company Nuclear Safety Advisory Letter This is a notification of a recently identified potential safety issue pertaining to basic components supplied by Westinghouse. This information is being provided to you so that a review of this issue can be conducted by you to determine if any action is required. P.O. Box 355, Pittsburgh, PA Subjectý-- -AX]AL-BU-RNUP- SHAPE--R-A-C-T-1-V-[T---BIAS Number-: NSAL--W Basic Component: Spent Fuel Criticality Anialysis [Date: November 2, 2000 Plants: Beaver Valley 2, Comanche Peak I & 2, Turkey Point 3 & 4, Maanshan I & 2, Indian Point 3, Millstone 3, V. C. Summer, Braidwood 1 & 2, Farley I & 2, Prairie Island I & 2, South Texas I & 2, Vogtle I & 2, Krsko, North Anna 1 & 2 Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR (a) Yes LI No [ Transfer of Information Pursuant to 10 CFR (b) Yes L1 Advisory Information Pursuant to 10 CFR (d)(2) Yes El References: WCAP NP-A, Revision ], "Westinghouse Spent Fuel Rack Criticality Analysis Methodology," Newmyer, W D., November SUMMARY The Westinghouse methodology for determining spent fuel criticality allows a criticality credit by considering the reactivity decrease associated with fuel depletion (called bumup credit reactivity equivalencing). The methodology relies on two-dimensional (2D) radial calculations using PHOENIX-P. To account for axial, or three-dimensional (3D), burnup effects, a reactivity "bias" was identified on a generic basis in the referenced WCAP. This methodology has been used to establish plant technical specification limits. Recent calculations, using the Westinghouse methodology, have shown that the calculated axial burnup bias could be non-conservative. Close examination of the methodology also revealed several areas where additional credits could be technically justified. In some cases, these credits were not assumed as part of the methodology topical but are acknowledged, relevant factors as to the potential condition of the spent fuel pool. Information is provided within this advisory letter for plant-specific evaluations necessary to conform to the guidance of US NRC Generic Letter Revision 1, if required. The reactivity bias penalty due to the non-conservatism was calculated along with the quantification of the technically justifiable credits on a generic basis. For potentially affected plants, the results show the net effect to be no more limiting than results calculated using the WCA-P methodology. That is, the current technical specification limits on spent fuel pool loading configurations for potentially affected plants are technically acceptable and the curves in the technical specification remain valid. On this basis, Westinghouse has identified no compensatory actions on the part of the potentially affected plants. Plants that use the WCAP NP-A burnup credit methodology, in whole or part, should review their analysis method to determine if the non-conservative axial-bias applies to them. Additional information, if required, may be obtained from the originator. Telephone Originator(s): D. A. Lindgren Regulatory and Licensing Engineering W. H. Slagle A'll Core Analysis B H. A. Sepp Regulatory and Licensing Engineering

109 ATTACHMENT 1 NSAL Page 2 of 8 ISSUE DESCRIPTION The Westinghouse methodology for spent fuel criticality is described in Reference 1, and has been approved by the NRC. The methodology includes a technique where credit can be taken in criticality analyses by considering the reactivity decrease associated with fuel depletion (called burnup credit reactivity equivalencing). The calculations for burnup credit reactivity equivalencing are done on a radial, two-dimensional (2D) basis with the PHOENIX-P code. Inherent in a 2D treatment for this calculation is a uniform axial burnup distribution. To account for the varying burnup and reactivity axially along the assembly, that is, the threedihmesional (3D)buriuip effect, a-bia-s term-h-ad- beeni defined inreference 1-us ing the PHOENIX-_P and ANC codes. An issue was identified in a calculation of burnup credit for the spent fuel rack criticality analysis by a utility owner for its plant of Westinghouse design. Using a different methodology and set of assumptions, and alternate calculation codes, the calculation suggested a larger axial bias term with a lower burnup application limit (burnup at which the "bias" results in more severe reactivity results) than that calculated by Westinghouse. The NRC was notified of these preliminary results in March 2000, by the utility, prompting an investigation by Westinghouse with respect to the reported sensitivity. The listed plants are those for which Westinghouse Electric has performed the spent fuel pool criticality analysis or those plants for which Westinghouse knows that the WCAP NP-A methodology is used to determine axial bias. Since WCAP NP-A was developed as part of a Westinghouse Owners Group project, it was available for use by any participating utility. LICENSING BASIS There are two general design criteria (10 CFR Part 50 Appendix A) that are relevant to the analysis of spent fuel pool criticality. General Design Criterion 61 (Fuel Storage and Handling and Radioactivity Control) provides that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity, shall be designed to assure adequate safety under normal and postulated accident conditions. General Design Criterion 62 (Prevention of Criticality in Fuel Storage and Handling) provides that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. TECHNICAL EVALUATION An investigation was conducted which focused on various aspects of the spent fuel pool criticality analysis. Both the methods and assumptions used in the analysis were reviewed. The first part of the investigation centered on a detailed review of the methods and assumptions used in the Westinghouse analyses and those used by others in the nuclear industry. As a result of this part of the investigation, the axial bias non-conservatism was verified and several conservatisms were identified and documented.

110 ATTACHMENT 1 NSAL Page 3 of 8 This part of the investigation was conducted by Westinghouse personnel and other nuclear industry individuals. The magnitude of non-conservatism varies from plant to plant. It also varies among configurations. The typical range of values of the non-conservatism is 0 to delta-k based on the configuration and burnup. For some checkerboard type loading patterns the effect of the non-conservatism on criticality calculations may be as low as zero. The maximum value for non-conservatism in the axial bias is typically for configurations of the highest required burnup. The second part of the investigation was a generic evaluation of previous analyses performed by Westinhiolus-e,- for -v--aou-s-pla-ets,-wicih accoun-te d- for the findings of- the detailed review conducted in the first part of the investigation. From these evaluations, certain conservatisms in the methodology also were confirmed which would tend to offset the non-conservative nature of the bias, if they were to be factored into the previous plant-specific analysis. The conclusion of the evaluations was that the credit for the overall conservatisms identified in the survey of plant analyses is sufficient to offset the effects of the revised axial burnup bias identified above. It should be noted that for some plants, which credit soluble boron in defining the bumup credit, the actual amount of boron needed to meet the kef- < 0.95 requirement is known and was properly allocated. However, additional available boron was not credited in the evaluations presented herein nor was soluble boron credit assumed for plants that are not licensed for soluble boron credit. EVALUATION OF CONSERVATIVE CREDITS The evaluations were conducted for plants for which the Westinghouse methodology for determining spent fuel criticality had been applied. Generic conservatisms, applicable to these plant analyses, were determined which showed that the plant analyses would be acceptable on a net basis. That is, the potential non conservatism of the axial burnup bias would be less than the demonstrated generic conservatisms identified. It should be noted that in evaluating the previous plant-specific analyses, not all of the generic conservatisms have been used and not to the same degree. The conservatisms and the calculation of the credits are described below. The values stated for the credits are typical values and are subject to plant specific variation. These generic values may not add up to the delta-k that was previously identified as the upper range of the non-conservatism. Discrete Lattice Single Rack Cell Assumpibon Westinghouse has typically represented each spent fuel pool with an infinitely repeating single cell assumption. This assumption, by definition, produces zero leakage in the horizontal direction of the spent fuel pool. Also, because of this single cell assumption, Westinghouse has modeled all of the inventory in the spent fuel pool with a single assembly description. The purpose of the evaluation is to quantify the reactivity associated with the leakage out of modules (which is a collection of fuel assemblies separated by varying amounts of water) and the reactivity effect associated with a more realistic, and yet conservative, description of the inventory in the spent fuel pool. The analysis directly simulated the reactivity effects of leakage out of a module and the "global" mixing of different assembly burnups within a typical spent fuel pool. The leakage out of a typically sized module (lox10 arrangement of fuel assemblies) was calculated for two different intra-module (between module) gap sizes: 2.0 cm (0.79 inch) and 5.08 cm (2 inches). The calculated leakage for the 2 cm gap between modules is delta-k. The calculated leakage for the 5.08 cm gap between modules is delta-k.

111 ATTACHMENT 1 NSAL Page 4 of 8 The assumption that all fuel assemblies are discharged into the spent fuel pool with a single limiting axial burnup profile is very conservative. In reality, the fuel assemblies are discharged into the spent fuel pool with axial burnup profiles that would produce lower reactivity results. However, it is reasonable to assume that the magnitude of this effect is approximately equal to the reactivity associated with the mixing of modules loaded with varying assembly bumups. The reactivity effect associated with the mixing of different assembly burnups within a spent fuel pool was approximated. This was accomplished by checker-boarding modules with different assembly burnups. The minimum calculated reactivity worth of this effect is delta-k and the maximum calculated reactivity worth is delta-k. The average of these two values, delta-k, would r asnbf~ reprsn te spectrum of assem~ urnups expected within a typical spent fuel pool. It is expected that a checkerboard of differing burnup assemblies within a rack module would give at least twice the benefit. Therefore, the total reactivity effect of varying the axial burnup shapes within a module and varying the spectrum of assembly burnups within the spent fuel pool is calculated to be delta-k. The total combination of the expected leakage out of the single cell assumption along with the mixing of axial burnup profiles and assembly burnups within the spent fuel pool is delta-k (e.g., delta-k delta-k). Thus, it has been conservatively reduced to delta-k. Presence Of Samarium And Fission Product Buildup In the current methodology, no credit for samarium and fission products is assumed. The typical minimum cooling time is 100 hours after shutdown before the fuel movement can be performed. One hundred (100) hours will cover most plants for a generic analysis. The one plant that is an exception (42 hours cooling) does not take samarium credit. Typical fuel enrichments for reload cores range from 3 to 5 w/o. So the data for the 3 and 5 w/o values are used. This is deemed conservative because the data trends seem to credit samarium less with increased enrichment. The credits for a range of burnups are used to develop the average value. Assemblies with only 30,000 MWD/MTU burnup or lower are typically not of concern for axial burnup bias. The bias for 30,OOOMWD/MTU assemblies is not substantial. For assemblies that are not at full power, for example periphery assemblies, a penalty must be applied in final accounting. A part-power factor of 0.7 is taken and used. The typical samarium credit for the buildup of samarium and fission products after shutdown results in a negative reactivity credit of approximately delta-k. Decay Time Credit Spent fuel decay time credit results from the radioactive decay of isotopes in the spent fuel to daughter isotopes. This decay results in reduced reactivity. Credit is taken only for the decay of actinides, mainly the decay of Pu-241 (fissile isotope) to Am-241 (poison absorber). Westinghouse has evaluated up to 5 years of decay time credit in evaluation of previous analyses for all cell configurations since this configuration has a large axial burnup bias. No decay time credit was necessary for checkerboard type configurations since this configuration has negligible axial burnup bias.

112 ATTACHMENT 1 NSAL Page 5 of 8 Pool Leakaqe In the current methodology, the storage cells are assumed to be infinite in the lateral (x and y) direction. In the actual storage pool, leakage occurs between the rack module and into the pool wall. The gap between the racks and the sides of the pool was set at 2 inches and larger. A concrete boundary of 24 inches was used. The results for the various gap sizes at the wall are virtually identical. The model also considers no absorber panels on external faces of periphery cells. Based on the evaluations of previous analyses, it has been determined that delta-k credit can be taken on a generic basis when accounting for the leakage between storage rack modules and the pool Wall. -f additionalcredit is-needed, a-plant specifi- en-gineern evaluation can be made to raise the credit to as high as delta-k. Boron LetdoWn Curve For Hot Full PoWer Depletion In the current methodology, the fuel assembly depletion is performed with a constantly high value of soluble boron (e.g., a value of ppm is used for bumup from 0 to 60,000 MWD/MTU). In actual operation, the soluble boron decreases during the cycle (e.g., typically a high boron value at the beginning-of-cycle, and a near zero value at the end-of-cycle). The lower cycle average boron value, for actual operations, results in a softer neutron spectrum, and makes the fuel assemblies less reactive with burmup due to the smaller buildup of plutonium. For the generic evaluation of previous analyses described earlier, a bounding boron letdown curve was assumed (e.g., 1500 ppm boron at beginning-of-cycle to 0 ppm at end-of-cycle) that also included burnable absorbers; therefore, the assemblies in the core will be less reactive. Existinq Delta To The KefLimit This is the difference between the kff limit of 0.95 (for no soluble boron credit) or 1.00 (for soluble boron credit) and the calculated value of kff determined on a 95/95 basis. Grid And Sleeve Credit Under the current methodology, no credit is taken for the presence of grids and sleeves. This credit is determined to be in the range of delta-k to delta-k, depending on fuel type. A portion of this credit is reserved for another issue, leaving to delta-k. The following three credits are grouped since the tolerance uncertainties for enrichment, density, dishing and others, are handled in the current methodology through statistical convolution. Therefore, it is not practical to separate out individual impacts and determine a credit. This is specifically true since the magnitude of the following three credits are plant-specific and a generic value can not be provided as representative for all plants. Enrichment Tolerance In the current methodology, the standard DOE tolerance is ± 0.05 w/o U235 about the allowable enrichment for fresh fuel with no burnup. The allowable initial enrichment in the base methodology is usually low (less than 2.0 w/o) for storage in all cells. This results in a rather large uncertainty

113 ATTACHMENT 1 NSAL Page 6 of 8 from a reactivity standpoint. Note: the assembly enrichment in reload cores is usually in the range of 3.0 to 5.0 w/o. The enrichment tolerance uncertainty for high bumup fuel at a higher enrichment of up to 5.0 w/o U235 is significantly smaller. Density Tolerance In the current methodology, a ± 2.0% variation about the nominal U0 2 theoretical density of 95 to 96% is used. The specification for the maximum theoretical density at the Westinghouse fuel manufacturing site is 96.5% on a fuel assembly average basis. Therefore, the density tolerance on the positive side should be only +0.5 to +1.5%. _This results-in a-lower-density reactivity- tolrane-e nimcerftain compared to the current methodology. * Dishing Tolerance In the current methodology, a 0% pellet effective dishing is assumed for the dishing tolerance uncertainty. This results in a measureable dishing reactivity tolerance value. Pellets are actually manufactured with a chamfer and dishing. Therefore, a conservative approach is to take credit only for 50% of the nominal dishing. This results in a significant reduction in the dishing reactivity tolerance value. SUMMARY To reiterate, the conclusions drawn from the evaluation are that the credits for the overall conservatisms identified in the survey of plants are sufficient to offset the effect of the revised axial burnup bias. Analyses based on the referenced WCAP that have not yet been completed or that were not performed by Westinghouse may employ a similar approach to demonstrate the conservatism of the results. Alternately, the axial burnup bias may be specifically addressed in the analysis to ensure that a conservative approach is used. EFFECT ON DESIGN BASIS AND DESIGN REQUIREMENTS Adequate safety under normal and postulated accident conditions for fuel storage and handling (GDC 61) is provided by the structural integrity operational performance of the spent fuel pool, spent fuel racks, and the spent fuel pool cooling system. The method of analyzing spent fuel pool criticality does not affect structural integrity or operational performance. Preventing criticality in the fuel storage and handling system (GDC 62) is provided by maintaining keff less than or equal to 0.95, at a 95% probability, 95% confidence level when accounting for the presence of boron or kff less than or equal to 1.00, at a 95% probability, 95% confidence level when not accounting for any boron presence. The evaluation of the net effect of the non-conservative axial burnup bias and the available credit for conservatisms in the WCAP NP-A methodology demonstrates that this requirement continues to be satisfied. The interface between the fuel assembly and the fuel handling equipment and storage racks is not altered. The design requirements for the fuel storage racks are not altered. The spent fuel handling tools do not have to be altered. The fuel handling procedures are not affected.

114 ATTACHMENT I NSAL Page 7 of 8 The fuel assemblies, fuel rods, evaluation of criticality in the reactor vessel during refueling, and performance of the reactor coolant and supporting systems are not affected by the non-conservatism in the spent fuel criticality methodology. PLANT TECHNICAL SPECIFICATIONS The plant Technical Specifications typically includes limits on spent fuel pool rack These loading limits configurations. provide a margin of safety with respect to criticality in the spent account fuel storage for fuel racks. assembly The enrichment limits and burnup. The technical specification are limits based for on the the WCAPaffected I NPA plants methodology- Removing some- of the with eonservatisms-fro-m-the-- the revised axial bias and -a-ftyis accounting for the additional identified conservatisms specification shows limits that remain the technical valid. Fuel assemblies do not have to be moved. The limit on reactivity in the spent fuel pool is not reduced. The use of the conservatisms in the calculation of the limits on placement of the spent fuel represents a change to the evaluation that supports the normal plant configuration. The FSAR discussion of criticality in the spent fuel pool and the description of the basis of the technical specification should be reviewed for impact. ASSESSMENT OF SAFETY SIGNIFICANCE Evaluations factoring in the axial bias correction and using identified conservatisms in the analysis, and acknowledgment of actual conditions demonstrate that kff remains less than or equal to the limit for potentially affected plants. Therefore, the current spent fuel pool configurations for the potentially affected plants are technically acceptable and continue to provide a geometrically safe configuration. NRC AWARENESS/REPORTABILITY CONSIDERATIONS The utility that identified this issue informed the NRC via a Licensee Event Report (Event Number 36748, 03/02/2000). Westinghouse has discussed this issue with the NRC staff with regards to the investigation, subsequent results, and notification plans. The evaluation of the credits for conservatism and the review of the technical specification basis outlined above demonstrate that k-ff remains less than There is no loss of safety function such that there is a reduction in the degree of protection provided to public health and safety. Therefore, this issue does not constitute a Substantial Safety Hazard as defined in 10 CFR Part 21 and is not reportable as a Part 21 item. RECOMMENDED ACTIONS The current spent fuel pool technical specification loading configurations for the potentially affected plants are technically acceptable based on the evaluations described herein. Regulatory and licensing commitments may require additional licensing activities. These may include a review of operation using the guidance of Generic Letter 91-18, Revision 1, preparation of a 10 CFR evaluation, or updating the basis of the technical specifications. The key assumptions used in the Westinghouse evaluations have been delineated herein.

115 ATTACHMENT 1 NSAL Page 8 of 8 Plants that use the WCAP NP-A burnup credit methodology in whole or part should review their analysis method to determine if the non-conservative axial-bias applies to them. On a plant-specific basis, the non-conservative axial bias calculations should be reviewed to determine if this condition represents a nonconforning or degraded condition as defined in US NRC Generic Letter 91-18, Revision 1. The documentation and evaluation requirements needed to address the guidance of Generic Letter 91-18, Revision 1 is determined on a plant specific basis. WCAP NP-A may be considered part of the plant design or licensing basis. The actions required to confirm the technical specification limits on spent fuel loading configurations is also a plant-specific effort. Evaluations performed to_ address- the,- guidance -of-generic-letter- 91-1&;-- Revision assessment 1 -mayof includecontinued ani operation. The items that are considered appropriate continued in the consideration operation include of conservatisms and margins. The conservatisms available analysis in the criticality are outlined above. The result that the credit for the conservatisms is sufficient to offset the effect the of revised axial bias may be used as the basis for continued operation. The evaluations that supported preparation of this advisory letter have determined that compensatory actions are not required. The fuel currently stored in the spent fuel pool does not have to be moved. The fuel assemblies of approved designs in the current or future cores may be stored in the spent fuel rack using existing limits.

116 0 Westinghouse Electric Company, LLC Box 355 Pittsburgh Pennsylvania June 12, 2001 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC CAW Attention: Mr. Samuel J. Collins APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE Subject: "Westinghouse Proprietary Information Related to Axial Burn-up Bias" (Proprietary), June 2001 Dear Mr. Collins: The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section of the Commission's regulations. Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Carolina Power and Light. Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW and should be addressed to the undersigned. Very truly yours, H.A SIeppA1anagr Regulatory and Licensing Engineering Enclosures cc: S. Bloom/NRR/OWFN/DRPW/PDIV2 (Rockville. MD) 1L P:DATA/DOCUMENTS/

117 CAW AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA: ss COUNTY OF ALLEGHENY: Before me, the undersigned authority, personally appeared Henry A. Sepp, who, being by me duly sworn according to law, deposes and says that lie is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief: L 4. Henry A. Sepp.vanager A "Regulatory' and Licensing Engineering Swo'Mo* and subscribed before me this. day of,2001 Notarial Seal Patricia L. Crown, Notary Public Monroeville Boro, Allegheny County My Commission Expires Feb. 7, 2005 Member, Pennsylvania Association of Notaries Notary Public P:DATA/DOCUMENTS/

118 P:DAFA./DOCUMENTS/ -2- CAW (1) I am Manager, Regulatory and Licensing Engineering, in the Nuclear Services of the Westinghouse Electric Company LLC ("Westinghouse"), and as such. I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse. (2) I am making this Affidavit in conformance with the provisions of OCFR Section of the Cormnission's regulations and in conjunction with the Westinghouse Application for Withholding accompanying this Affidavit. (3) I have personal knowledge of the criteria and procedures utilized by the Westinghouse Electric Company LLC in designating information as a trade secret, privileged or as confidential commercial or financial information. (4) Pursuant to the provisions of paragraph (b)(4) of Section of the Commission's regulations. the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld. (i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required. Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

119 -3- CAW (a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies. (b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability. (c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture., shipment, installation, assurance of quality., or licensing a similar product. (d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers. (e) It reveals aspects of past. present. or future Westinghousc or customer funded development plans and programs of potential commercial value to Westinghouse. (f) It contains patentable ideas, for which patent protection may be desirable. There are sound policy reasons behind the Westinghouse system which include the following: (a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore., withheld from disclosure to protect the Westinghouse competitive position. (b) It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information. P:DATA/DOCUMENTS/

120 -4- CAW (c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense. (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage. (e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries. (f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage. (iii) The information is being transmitted to the Commission in confidence and, uinder the provisions of IOCFR Section 2.790, it is to be received in confidence by the Comnmission. (iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief. (v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Westinghouse Proprietary Information Related to Axial Burn-up Bias" (Proprietary), June 2001 for the Comanche Peak Steam Electric Station, being transmitted by TXU Electric Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk, Attention: Mr. Samuel J. Collins. The proprietary information as submitted for use by TXU Electric Company for the Comanche Peak Steam Electric Station is expected to be applicable in P:DATA/DOCUMENTS/

121 Pf-TATA/DflOCI TMNTS/ CAW other licensee submittals in response to certain NRC requirements for justification of Axial Burn-up Bias. This information is part of that which will enable Westinghouse to: (a) Provide documentation of axial bum-up bias penalty and credits related to the spent fuel pool criticality analysis. Further this information has substantial commercial value as follows: (a) Westinghouse could sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation. (b) Westinghouse can sell support and defense of the technology to its customers in the licensing process. Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar licensing support documentation and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information. The development of the technology described in part by the information is the result of applying the results of many\ years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

122 -6- CAW In order for competitors of Westinghouse to duplicate this information, similar design programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing analytical methods and performing tests. Further the deponent sayeth not. P:DATA/DOCUM4ENTS/

123 Proprietary Information Notice For NRC Transmittal Letter Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. In order to conform to the requirements of 10 CFR of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

124 Copyright Notice For NRC Transmittal Letter The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

125 ENCLOSURE 6 to TXX Westinghouse Letter logged WPT from J. S. Wyble to C. L. Terry dated December 19, 2000 Attachment 2 to NSAL (Non-Proprietary)

126 Westinghouse Proprietary Information Related to Axial Bum-up Bias, June 2001 Westinghouse Non-proprietary Class 3 Westinghouse Electric Company Nuclear Services Box 158 Madison, Pennsylvania December 19, 2000 WPT Mr. C. L. Terry, Senior Vice President & Principal Nuclear Officer T-xU Electric Company P.O. Box 1002 Glen Rose, TX Attention: VETIP Coordinator No Response Required Ref 1) WPT Ref 2) WPT Rev 1 Dear Mr. Terry: TXU Electric Company Comanche Peak Steam Electric Station Unit Number 1 & 2 NSAL , Axial Bumup Shape Reactivity Bias The subject Nuclear Safety Advisory Letter regarding axial bumup shape reactivity basis was issued via Ref (1). Per TXUElectric request, we have revised Attachment 2 of the NSAL to Revision 2, adding an additional clarifying sentence to the first page of Attachment 2. The revised version of Attachment 2 is provided with this letter. If you have any questions concerning this Advisory Letter, please contact William Slagle at either or slaglewh( iwestinghouse.com, or Donald Lindgren at or lindgl da@westinghouse.com. Very truly yours, Attachment 2 (Revision 2) J S. o. Wyble Customer Projects Manager

127 Westinghouse Proprietary Information Related to Axial Bum-up Bias, June 2001 Comanche Peak Units I & 2 Spent Fuel Pool Region & Configuration: 2, All Cell Axial Bumup Bias Penalty: 4359 Summary of Credits: Presence of samarium and fission product buildup credit Discrete lattice single rack cell assumption credit 2 Boron letdown curve for HFP depletion credit Enrichment, density, dishing tolerance credit Existing delta to the kfr limit Grid and sleeve credit Pool leakage credit Decay time credit - 5 WCAP NP-A axial burnup bias credit _.j Net Balance: 47 Spent Fuel Pool Region & Configuration: 2, 3 of 4 Axial Bumup Bias Penalty: 2091 Summary of Credits: Presence of samarium and fission product buildup credit OC I Discrete lattice single rack cell assumption credit 2 Boron letdown curve for HFP depletion credit 3 Enrichment, density, dishing tolerance credit 4 Existing delta to the kdfr limit s Grid and sleeve credit 6 Pool leakage credit Decay time credit WCAP NP-A axial burnup bias credit L Net Balance: 463 There is no axial burnup bias penalty associated with the Region 2, 2 of 4 configuration since the burnup associated with this configuration does not result in a penalty. Note: AD units are xle AK

128 Westinghouse Proprietary Information Related to Axial Burn-up Bias, June 2001 Westinghouse Assessment of Credits The following discussion is provided for utility use in assessing the licensing position of the credits identified (e.g., allowed by the topical, not allowed by the topical, or not discussed in the topical). This assessment is Westinghouse's perspective of the licensing position on the credits. Utilities will have to determine whether they agree with Westinghouse's perspective or establish their own position. On page 4 of the SER, item 9, it states that "no amount of fission product material is modeled in the fuel assembly". This is an input assumption that was noted by the NRC as "... tend to maximize the rack reactivity and are, therefore, appropriately conservative and acceptable". However, it is also noted on page 6 of the SER, sixth paragraph, that the staff does not consider it a requirement to take no credit for fission product poison material in doing the criticality analysis. Therefore, it is appropriate to consider that samarium and fission product buildup credit is acceptable and within the bounds of the WCAP as approved by the NRC. On page 11 of the SER, item 1, it states that "if axial and planar variations of fuel assembly characteristics are present, they should be explicitly addressed, including the locations of burnable absorber rods". Since the origi al analysis was done assuming an infinite lattice, which is conservative, it is also acceptable to do discrete modeling of the assembly. Therefore, it is appropriate to consider that discrete lattice single rack cell assumption credit is acceptable and within the bounds of the WCAP as approved by the NRC. On page 8 of the topical, Section 4.1, second paragraph, it is noted that "a conservatively high soluble boron letdown curve is chosen to enhance the buildup of plutonium thus making the fuel assembly more reactive when stored in the spent fuel storage racks. The SER only states that appropriate fuel depletion be accounted for in the analysis. Since it was identified that a conservatively high soluble boron letdown curve would be used, then the use of a flat peak boron concentration during the depletion is an excessive conservatism that can be reduced to that allowed in the WCAP. Therefore, it is appropriate to consider that boron letdown curve for HFP depletion credit is acceptable and within the bounds of the WCAP as approved by the NRC. 4 On pages 3, 4 and 5 of the SER, it is stated that nominal values are to be used with the appropriate tolerances accounted for in the analysis. As noted in the NSAL, the previous analysis has assumed conservative values for these parameters and their tolerances. The rationale for accounting for the more realistic tolerances is also specified in the NSAL. Since the SER identifies allowances for the enrichment, density and dishing fraction tolerances, then accounting for these allowances can be used to obtain a revised uncertainty bias term (B )., Therefore, it is appropriate to account for the allowances and the Enrichment, density, dishing tolerance credit is acceptable and within the bounds of the WCAP as approved by the NRC. The existing delta to the kff limit is the difference between the kif limit of 0.95 (for no soluble boron credit) or 1.00 (for soluble boron credit) and the calculated value of kdr determined on a 95/95 basis. This is margin between the analysis results and the limit and does not need to be addressed in the topical report On page 4 of the SER, item 7, it states that "no amount of material from spacer grids or spacer sleeves is modeled in the fuel assembly". This is an input assumption that was noted by the NRC as "... tend to maximize the rack reactivity and are, therefore, appropriately conservative and acceptable". It should be noted that the NRC did not state that these assumptions are required. In fact, as noted in item 1 above for the fission product assumption, the NRC later stated in the SER that "the staff does not consider this to be a requirement". Since the NRC did not state that the assumptions are requirements in Section 3.2 of the SER and in one particular case (i.e., the fission products) it was stated that the staff does not consider it a requirement, it could be construed that none of the assumptions listed in Section 3.2 of the SER are requirements. However, it is noted that there is no additional discussions on this assumption. Therefore, from a prudence standpoint, Westinghouse would assume that grid and sleeves should not be modeled as a credit if one is strictly abiding by what is allowed within the topical report. Note: All units are xl0s AK