HBL-14-003, Revision 10 to the Defueled Safety Analysis Report and Revision 31 to Quality Assurance Plan

Revision 10 to the Defueled Safety Analysis Report and Revision 31 to Quality Assurance Plan

ML14035A514
Person / Time
Site:	Humboldt Bay
Issue date:	02/04/2014
From:	Halpin E Pacific Gas & Electric Co
To:	Document Control Desk, NRC/FSME
References
HBL-14-003
Download: ML14035A514 (91)

Text

Pacific Gas and Electric Company Edward D. Halpin Diablo Canyon Power Plant Senior Vice President Mail Code 104/6 Nuclear Generation & Chief Nuclear Officer P. 0. Box 56 Avila Beach, CA 93424 805.545.4100 E-Mail: E1HB@pge.com February 04, 2014 PG&E Letter HBL-14-003 U.S. Nuclear Regulatory Commission 10 CFR 50.71 ATTN: Document Control Desk 10 CFR 50.54 Washington, DC 20555-0001 Docket No. 50-133, DPR-7 Humboldt Bay Power Plant, Unit 3 Revision 10 to the Defueled Safety Analysis Report and Revision 31 to the Humboldt Bay Quality Assurance Plan

Dear Commissioners and Staff:

Pursuant to 10 CFR 50.71(e), Pacific Gas and Electric Company (PG&E) is submitting the Humboldt Bay Power Plant (HBPP), Unit 3 Defueled Safety Analysis Report (DSAR), Revision 10, in accordance with 10 CFR 50.4(b )(6). The changes contained in Revision 10 of the DSAR include changes made under the provisions of 10 CFR 50.59 since the DSAR, Revision 9, was submitted to the NRC in PG&E Letter HBL-12-003, dated February 21, 2012, and reflect the current plant configuration and organization through November 30, 2013. contains Revision 10 of the DSAR in its entirety. The DSAR changes were made under the provisions of 10 CFR 50.59 and have not been previously submitted to the NRC. Revision bars are not shown in Revision 10 of the DSAR due to the extensive changes made since Revision 9.

PG&E is also submitting Revision 31 of the Humboldt Bay Quality Assurance Plan (QAP) in accordance with 10 CFR 50.54(a)(3). Enclosure 2 contains Revision 31 to the Humboldt Bay QAP in its entirety. Revision bars are not shown in the enclosed version of Humboldt Bay QAP Revision 31 due to the extensive changes made since Revision 28 was submitted to the NRC in PG&E Letter HBL-12-003, dated March 3, 2012.

There are no regulatory commitments (as defined by NEI 99-04) in this letter.

If you have any questions or require additional information, please contact Mr. David Sokolsky at (415) 973-5024.

Document Control Desk PG&E Letter HBL-14-003 February 04, 2014 Page 2 I state under penalty of perjury that the foregoing is true and correct.

Executed on February 04, 2014.

Sincerely,

~/~~

Edward D. Halpin Senior Vice President and Chief Nuclear Officer Enclosures cc: HBPP Humboldt Distribution cc/enc: Marc L. Dapas, NRC Region IV Administrator John B. Hickman, NRC Gonzalo Perez, California Department of Public Healt

Enclosure 1 PG&E Letter HBL-14-003 Humboldt Bay Power Plant, Unit 3 Defueled Safety Analysis Report, Revision 10

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS Section CHAPTER 1. INTRODUCTION AND

1.1 INTRODUCTION

1-1 1.2 SITE CHARACTERISTICS 1-2 1.3 QUALITY STANDARDS 1-2 1.4 FIRE PROTECTION 1-2 1.5 CONTROL OF RELEASES OF RADIOACTIVE MATERIALS 1-2 1.6 PERSONNEL SELECTION AND TRAINING 1-3 1.7 REPORTING OF OPERATING INFORMATION 1-3 1.8 MEASURING AND REPORT OF EFFLUENTS FROM NUCLEAR POWER PLANTS 1-3 1.9 IDENTIFICATION OF AGENTS AND CONTRACTORS 1-4 1.10 CONCLUSIONS 1-4 CHAPTER 2. SITE AND ENVIRONMENT 2.1

SUMMARY

2-1 2.2 SITE AND ADJACENT AREAS 2-1 2.3 METEOROLOGY 2-1 2.4 HYDROLOGY 2-1 2.5 ENVIRONMENTAL MONITORING PROGRAM 2-2 CHAPTER3. STRUCTURES 3.1 REFUELING BUILDING 3-1 3.2 RADWASTE TREATMENT FACILITIES 3-1 REV10

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS Section CHAPTER 4. AUXILIARY SYSTEMS 4.1 GENERAL 4-1 4.2 VENTILATION SYSTEMS 4-1 4.3 WASTE DISPOSAL SYSTEMS 4-1 4.4 FIRE PROTECTION SYSTEMS 4-1 CHAPTER 5. RADIOACTIVE WASTE AND RADIATION PROTECTION 5.1 RADIONUCLIDE INVENTORY 5-1 5.2 LIQUID WASTE TREATMENT SYSTEMS 5-1 5.3 GASEOUS WASTE MANAGEMENT 5-1 5.4 SOLID WASTE MANAGEMENT 5-1 5.5 OFF-SITE DOSE CALCULATION MANUAL 5-1 5.5.1 Liquid Discharge Pathway 5-2 5.5.2 Gaseous Discharge Pathway 5-2 5.6 RADIATION PROTECTION PROGRAM 5-2 5.6.1 Organization 5-2 5.6.2 Equipment and Instrumentation 5-2 5.6.3 Radiation Protection Procedures 5-2 5.6.4 Personnel Exposure Reporting 5-2 CHAPTER 6. CONDUCT OF OPERATIONS 6.1 PLANT ORGANIZATION 6-1 6.2 PERSONNEL TRAINING 6-1 6.3 EMERGENCY PLAN 6-1 REV 10

DEFUELED SAFETY ANALYSIS REPORT TABLE OF CONTENTS Section 6.4 PLANT PROCEDURES 6-1 6.5 RECORDS 6-1 CHAPTER 7. SAFETY ANALYSIS 7.1 DECOMMISSIONING ACCIDENT ANALYSIS 7-1 Appendix A Implications of Decommissioning Accidents with Potential for A-i Radiological Impacts to the Environment iii REV 10

DSAR CHAPTER INTRODUCTION AND

1.1 INTRODUCTION

In 1984 PG&E submitted the Humboldt Bay Power Plant, Unit 3 (referred to as either HBPP, Unit 3, or the Humboldt Bay nuclear facility) SAFSTOR Decommissioning Plan (SOP) in support of the application to amend the HBPP Operating License to a Possession-Only License. As a result of the 1996 NRC decommissioning rule, the SOP was considered to be a Post-Shutdown Decommissioning Activities Report (PSDAR) because it contained information related to decommissioning activities. It was also considered to be a Final Safety Analysis Report (FSAR) because it contained information such as plant description, site characterization and accident analysis. In compliance with the 1996 NRC decommissioning rule, PG&E submitted a PSDAR In February 1998 to provide a general overview of proposed decommissioning activities. As a result, the SOP will focus on providing the type of information contained in an FSAR and will contain less information related to decommissioning activities. Thus, the SOP has been more appropriately renamed the Defueled Safety Analysis Report (DSAR).

In addition to the DSAR and PSDAR, PG&E has submitted other documents to the NRC in accordance with 10 CFR 50 that constitute the licensing basis for HBPP. These other documents include: (1) License Amendments, (2) revised Technical Specifications, (3)

Environmental Report, (4) Quality Assurance Plan, (5) Security Plan, (6) Emergency Plan.

This DSAR reflects the changes made following Humboldt Bay's(HB's) decision to decommission Unit 3 in June 1983, as well as License Condition 2.8.1 that allow Pacific Gas and Electric Company (PG&E) " ... to take actions necessary to decommission and decontaminate this facility." PG&E will update the DSAR in accordance with the methodology specified in 10 CFR 50.59 and 10 CFR 50.71(e), submitting updates at least every two years. Revisions to the DSAR are numbered, starting with Revision 1.

The purpose of the DSAR is to provide a Safety Analysis Report that provides assurance that based on the decommissioning activities and the administrative controls and programs in place during decommissioning, no undue risk to the public health and safety will occur during normal operation and postulated accident conditions.

The following provides a history of major plant operation and licensing-related actions:

1. Unit 3 achieved initial criticality on February 16, 1963, and began commercial operation in August 1963.

2. On July 2, 1976, Unit 3 was shut down for annual refueling and to conduct seismic modifications. Seismic and geologic studies were in progress.

3 In December 1980 it became apparent that the cost of completing required backfits might have made it uneconomical to restart the unit. Work was suspended at that time awaiting further guidance regarding backfitting requirements.

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4 In 1983, updated economic analyses indicated that restarting Unit 3 would probably not be economical, and in June 1983 PG&E announced its intention to decommission the unit.

5 On December 11, 2008, PG&E completed the transfer of spent fuel from the spent fuel pool into the ISFSI.

1.2 SITE CHARACTERISTICS The Humboldt Bay nuclear facility is located about four miles true southwest of the city of Eureka, Humboldt County, California, and consists of 142.9 acres of land.

Since there are no design basis accidents that could approach 10 CFR 100 limits, no Safety Features are necessary to keep public exposures below the 10 CFR 100 dose limits under worst case postulated accident conditions.

1.3 QUALITY STANDARDS The Humboldt Bay Quality Assurance Plan (HBQAP) describes the 10 CFR 50, Appendix B quality assurance program.

1.4 FIRE PROTECTION The Fire Prevention Program describes the overall fire protection program in place at HB.

DSAR Chapter 4 also discusses the fire protection system.

1.5 CONTROL OF RELEASES OF RADIOACTIVE MATERIALS The radioactive waste systems and radiological administrative control programs at Humboldt Bay effectively control the collection, segregation, processing, packaging, and discharge or disposal of radioactive solids and liquids. These systems and programs function to control releases to as low as is reasonably achievable levels and are designed to give reasonable assurance that the numerical guidelines defined in 10 CFR 50.34a, 50.36a, and 10 CFR 50, Appendix I, and contained in the Off-site Dose Calculation Manual (ODCM) will be met.

During dismantlement, all ventilation treatment systems have been removed, or are being removed, with the exception of the HEPA filter on the Refueling Building effluent release point. This release point and the HEPA filter will be shut down and removed once the Refueling Building Structure is prepared for demolition.

Interim storage buildings are provided for packaged low-level radioactive waste. Potential release pathways of radionuclides in particulate form are controlled and monitored in accordance with the ODCM.

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Liquid and solid wastes are normally processed in batches for off-site disposal. Gaseous effluents (particulates and tritium) could be released during dismantlement activities. During this time, samples will be taken from the Refueling Building ventilation system, or other temporary monitored release points in accordance with the ODCM.

1.6 PERSONNEL SELECTION AND TRAINING Plant personnel meet the minimum education and experience standards specified in ANSI N18.1-1971, "Standard for Selection and Training of Personnel for Nuclear Power Plants."

Retraining and replacement training meets or exceeds the requirements and recommendations of ANSI N 18.1-1971. The Radiation Protection Manager shall meet or exceed the requirements of Regulatory Guide 1.8, Revision 2, April 1987.

DSAR Chapter 6 discusses training programs for staff personnel. Retraining and replacement training and records of the qualifications, background, training, and retraining of each member of the plant organization are maintained in accordance with established programs.

In accordance with 10 CFR Part 19, personnel are provided radiological training commensurate with the hazards to which they may be exposed while working at HBPP.

1.7 REPORTING OF OPERATING INFORMATION The Humboldt Bay Quality Assurance Plan specifies the annual routine reporting requirements.

Personnel exposure and monitoring procedures have been established to ensure that each individual is provided with records of their exposure, and to safeguard against exceeding exposure limits in the plant. Reports are submitted as required.

A program for administrative and physical control of radioactive sources containing special nuclear material has been established to ensure that the proper reports are provided.

Reporting abnormal events in Licensee Event Reports and Special Reports for issues such as overexposure and excessive radiation levels, loss of special nuclear material, and accidents involving licensed material is performed by site staff in accordance with established procedures and applicable regulatory requirements. Established procedures ensure significant events are quickly brought to the attention of management and acted upon within the allowed reporting period.

Humboldt Bay also supplies a voluntary report under the Nuclear Energy Institute Ground Water Protection Initiative.

1.8 MEASURING AND REPORT OF EFFLUENT FROM NUCLEAR POWER PLANTS The measuring and reporting of radiological effluents from Humboldt Bay complies with the requirements of Regulatory Guide 1.21 and the ODCM.

Annual reports summarizing the quantity of radionuclides released from the site are submitted to the NRC in accordance with 10 CFR 50.36a(a)(2). This summary data is comprised of radioactive release information collected in accordance with the ODCM.

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1.9 IDENTIFICATION OF AGENTS AND CONTRACTORS PG&E is responsible for control of agents and contractors used during the decommissioning the Humboldt Bay facility.

1.10 CONCLUSIONS The personnel assembled to decommission Humboldt Bay facility are capable of performing their required project function. The health and safety of the public and plant personnel are among the primary concerns during decommissioning.

The Humboldt Bay facility can be decommissioned in a safe manner and PG&E is qualified to decommission this nuclear facility in accordance with all applicable laws and regulations and in a manner satisfactory to the NRC, the public, and itself.

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DSAR CHAPTER 2 - SITE AND ENVIRONMENT 2.1

SUMMARY

The Humboldt Bay nuclear facility lies in the Northern California Coast Ranges geomorphic province. This province consists of a system of longitudinal mountain ranges (2000 to 4000 foot elevations with occasional 6000 feet peaks) and valleys with a trend of N 30 degrees to 40 degrees W. Terrain of the site varies from submerged and low tidal land, protected by dikes and tide gates, to a high precipitous bluff along the southwestern boundary. Elevations range from approximately -3 feet to +75 feet based on a datum of the mean lower low water (MLLW) level.

Soils in the immediate vicinity of the site consist of sand and alluvial soil and strata of the Hookton and Carlotta sedimentary formations. These formations are primarily consolidated sands, gravels, and clays and conglomerates with good engineering properties.

Figure 2-1 shows the general location of the Humboldt Bay site. Figure 2-2 provides a more specific layout of the PG&E-owned land and a general layout of the site.

2.2 SITE AND ADJACENT AREAS The land in the general vicinity around the Humboldt Bay site is presently low density rural housing and is used primarily as grazing areas.

Within a 15-mile radius of the site there are five cities or communities (Eureka, Fortuna, Arcata, King Salmon, and Fields Landing). Activities in the area immediately surrounding the site are not expected to change extensively.

King Salmon Avenue runs along the west side of the site boundary and connects with US Highway 101 which runs along the southern boundary of the site. The Southern Pacific Railroad runs roughly parallel to Highway 101 but is out of service.

PG&E is operating Humboldt Bay Generating Station (HBGS), a* new dual fueled (natural gas and diesel) power plant, on the Humboldt Bay site adjacent to the south side of the facility.

The Humboldt Bay ISFSI is located on site approximately 600 feet northwest of the HBGS.

2.3 METEROLOGY The Humboldt Bay facility uses conservative default atmospheric dispersion and deposition factors in its evaluations of off-site gaseous releases to calculate the doses that result from normal gaseous effluent and abnormal or accident release conditions.

2.4 HYDROLOGY The surface runoff from the site is directed into drains eventually discharging into Humboldt Bay.* Outside the area served by the plant drain system, surface runoff drains into Buhne Slough, the natural drainage for the area, which drains into Humboldt Bay.

The nearest streams to the site are Salmon Creek and Elk River, which are within a mile south and north of the site, respectively, and which discharge into Humboldt Bay. These streams are used for watering livestock, but are not used as a potable water supply.

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The Mad River flows west approximately 13 -15 miles northeast of the site. The Ruth reservoir, the source of the city's water supply, is located on this river.

To the south, the Eel River discharges to the Pacific Ocean 8-10 miles from HB. This river is not used for potable water within 25 miles of HB.

Groundwater elevation in the area near the bay is similar to sea level and may be somewhat affected by tidal action. This elevation is approximately 12 feet below the plant floor elevation.

2.5 ENVIRONMENTAL MONITORING PROGRAM The environmental monitoring of the environment during decommissioning is delineated in the Off-site Dose Calculation Manual.

Results of the Humboldt Bay environmental monitoring are available to the State of California and federal agencies that have a direct interest and concern in these matters.

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DSAR CHAPTER 3 - STRUCTURES 3.1 REFUELING BUILDING The refueling building is not required to function except for ALARA considerations to minimize potential occupational personnel exposures and for safe storage of contaminated systems, structures and components until decommissioned. Once the majority of radioactivity is removed from within the Refueling Building, the structure will be removed.

3.2 RADWASTE TREATMENT FACILITIES The radwaste treatment building contains tankage and a liquid treatment system. A solid waste vault is available to temporarily store radioactive material. Other buildings are used to temporarily store low level radwaste packages awaiting shipment. As decommissioning progresses, these structures, systems and components will be removed from service and removed.

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DSAR CHAPTER 4 -AUXILIARY SYSTEMS 4.1 GENERAL The systems considered in this DSAR Chapter are:

1. Ventilation Systems

2. Waste Disposal Systems

3. Fire Protection Systems Design changes, modifications, and additions to existing systems and components are reviewed for compliance with the requirements of 10 CFR 50.59 "Changes, Tests, and Experiments."

4.2 VENTILATION SYSTEMS The HVAC system for the refueling building and radwaste treatment building contains HEPA filters and fans. The gaseous effluent (particulate) is sampled for radioactivity in accordance with the Offsite Dose Calculation Manual (ODCM). The exhaust fan default flow rate is listed in the ODCM. As decommissioning progresses, areas ventilated by the system will change and eventually the entire system will be decommissioned and removed.

4.3 WASTE DISPOSAL SYSTEMS The waste disposal systems in Unit 3 include the liquid waste collection system, the liquid waste treatment and disposal system. Collectively these systems control and dispose of all plant wastes that are normally or potentially contaminated with radioactive materials.

Permanent installed systems may be removed and temporary systems instituted to aid in decommissioning. Eventually all radioactive waste disposal systems will be removed.

4.4 FIRE PROTECTION SYSTEMS Active and passive design features are provided to detect, contain, and suppress fires. The Unit 3 fire protection program includes the following fire protection features:

1. Fire suppression systems

2. Fire detection and ala~m systems The fire protection plan describes the organizational requirements, design features, operational requirements, compensatory measures, testing requirements, and off-site fire fighting assistance coordination requirements that collectively define Unit 3's fire protection program. Fire loading, ignition sources, and combustible materials are significantly reduced and the possibility for a major fire is greatly diminished from the unit's operational phase. As decommissioning progresses, the requirements for fire suppression, detection and alarm systems will be removed.

Offsite firefighting capability is provided by the local fire department.

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DSAR CHAPTER 5- RADIOACTIVE WASTE AND RADIATION PROTECTION 5.1 RADIONUCLIDE INVENTORY Radionuclides exist on site in various locations that will eventually be eliminated during decommissioning. Examples of places radionuclides exist or existed include the reactor vessel reactor vessel internals (that contain the largest percentage of onsite radionuclide inventory), and corrosion films within various in-plant systems.

These radionuclide sources are not readily dispersible in their present condition but will become more readily dispersible during decontamination and dismantlement activities.

Although the remaining radioactive source term at the defueled Unit 3 reactor site has been greatly reduced by radioactive decay and spent fuel removal, there still exist significant quantities of transuranic contamination within plant systems. Due to the internal hazard risk to workers of this transuranic contamination, administrative and engineering controls will be utilized to limit airborne radioactivity exposure to the workers and general public during decontamination and dismantlement.

5.2 LIQUID WASTE TREATMENT SYSTEMS During decommissioning, the liquid treatment system consists of various remaining permanent and temporary components that are used to transport and process liquid as necessary before final treatment and disposal. The wastewater disposal system includes tanks, pumps, piping, and processing equipment used to process waste water prior to discharge off site. Eventually all needs for liquid radioactive waste processing will cease and the permanent and temporary systems removed.

5.3 GASEOUS WASTE MANAGEMENT The Refueling Building Stack ventilation exhaust treatment system contains a HEPA filter, which filters the exhaust for particulates prior to release into the Refueling Building Stack vent. An installed monitor continuously monitors and samples for radioactivity releases in accordance with the ODCM.

5.4 SOLID WASTE MANAGEMENT Solid radioactive wastes are collected and processed on or off site and shipped to a licensed and/or permitted burial site for disposal. Off-site secondary processors may be used for volume reduction or further processing prior to disposal. Solid wastes are packaged in containers which conform to Department of Transportation (DOT) requirements (49 CFR) for transport to a licensed disposal facility.

The total curie content and major radionuclide composition by waste type are reported in the Radioactive Effluent Release Report required pursuant to 10 CFR 50.36a.

5.5 OFF-SITE DOSE CALCULATION MANUAL The ODCM provides the information and methodologies used to evaluate the impact of radiological liquid and gaseous effluent discharged from the plant.

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The ODCM is used to demonstrate that the plant complies with the requirements of 40 CFR 190 and 10 CFR 20, and the dose guidelines of 10 CFR 50, Appendix I. Calculations for continuous airborne releases use default atmospheric diffusion coefficients and gaseous effluent flow rates. Exposures due to plant decommissioning are estimated by calculation methods specified in the ODCM.

5.5.1 Liquid Discharge Pathway The liquid discharge pathway is eventually to Humboldt Bay. This pathway will receive treated liquid radwaste until the treatment system components are removed or the applicable permits are terminated.

This pathway, is designed to receive liquid from the ground water treatment system (GWTS).

Water collected is sampled and analyzed for radioactive discharge from the GWTS.

5.5.2 Gaseous Discharge Pathway The principal discharge source of normal radioactive gaseous effluent is the Refueling Building exhaust pathway. Releases from this pathway have been and are expected to continue to be very small; well within the 10 CFR 50, Appendix I dose guidelines and 10 CFR 20 concentration limits.

5.6 RADIATION PROTECTION PROGRAM 5.6.1 Organization The HB Director has the overall responsibility for radiological protection for all onsite activities, including assurance that ALARA policies are carried out at the facility. The Radiation Protection Manager is responsible for monitoring radiological conditions within the Industrial Area and the surrounding environs and specifying radiation protection requirements for work activities. The HBPP organization and reporting chain are specified in the HBQAP.

5.6.2 Equipment and Instrumentation Personnel Protective Equipment Anti-contamination clothing is furnished and worn as necessary to protect personnel against contact with radioactive contamination. Radiological respiratory protective equipment is available for the protection of personnel against airborne radioactive contamination.

Radiation Protection Instrumentation A variety of instruments are used to cover the entire spectrum of radiation measurements at Humboldt Bay.

5.6.3 Radiation Protection Procedures The radiation protection program shall be organized to meet the requirements of 10 CFR 20.

5.6.4 Personnel Exposure Reporting An annual exposure monitoring report for individuals is submitted in accordance with 10 CFR 20.2206(b).

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DSAR CHAPTER 6- CONDUCT OF OPERATIONS 6.1 PLANT ORGANIZATION The details of the PG&E organization that oversees the activities at the Humboldt Bay nuclear facility are discussed in the HBQAP and in administrative procedures. The HB Director heads the on-site Humboldt Bc;1y facility organization and directs the activities of the functional on-site departments. The HB Director is responsible to the PG&E Senior Vice President, Chief Nuclear Officer, for the safe decommissioning of the Humboldt Bay nuclear facility.

6.2 PERSONNEL TRAINING Plant personnel are selected and trained for their assigned duties, with each department head being responsible for conducting department training that meets the applicable requirements and standards.

Radiological training is provided to all personnel commensurate to the hazards to which an individual will be exposed. Training is conducted and documented in accordance with the training program.

6.3 EMERGENCY PLAN The Emergency Plan in effect during decommissioning provides a description of the organization, equipment, and preparations made to enable appropriate and effective response to postulated emergency situations that may arise at Humboldt Bay nuclear facility.

The focus of concern for the Emergency Plan is the protection of plant personnel and the surrounding population.

6.4 PLANT PROCEDURES The HB Administrative Procedures (HBAPs) define and implement administrative requirements or activities involving inter-departmental processes and administrative responsibilities. Procedure hierarchy, preparation, review, approval, revision, and control are established in accordance with the applicable HBAPs.

The HBQAP has been approved separately by the NRC. The plan describes the 10 CFR 50, Appendix B required Quality Assurance Program applicable to the Humboldt Bay nuclear facility. Changes to the QA Plan are made and submitted to the NRC in accordance with 10 CFR 50.59 and 10 CFR 50. 54( a).

6.5 RECORDS Records are maintained as described in the HBQAP.

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DSAR CHAPTER 7 - SAFETY ANALYSIS 7.1 DECOMMISSIONING ACCIDENT ANALYSIS Implications of decommissioning accidents with potential for radiological impacts to the environment are described in Appendix A.

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APPENDIX A IMPLICATIONS OF DECOMMISSIONING ACCIDENTS WITH POTENTIAL FOR RADIOLOGICAL IMPACTS TO THE ENVIRONMENT REV10 A-i

CONTENTS Section Page 1.0 Introduction and Background A-1 1.1 Dry Active Waste Fire A-2 1.2 Explosion of Liquid Propane Gas (LPG) Powered A-4 Equipment 1.3 Vacuum Filter Bag Rupture A-4 1.4 Contamination Control Envelope Rupture A-6 1.5 Oxyacetylene Explosion A-7 1.6 Filter Damage from a Blasting Surge A-9 1.7 Detonation of Unused Explosives A-10 1.8 Minor Transportation Accident A-11 1.9 Severe Transportation Accident A-12 1.10 HEPA Filter Fire A-12 1.11 Loss of HEPA Filtration A-14 2.0 References A-15 REV 10 A-ii

APPENDIX A IMPLICATIONS OF DECOMMISSIONING ACCIDENTS WITH POTENTIAL FOR RADIOLOGICAL IMPACTS TO THE ENVIRONMENT 1.0 Introduction and Background NUREG-0586, "Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," Supplement 1," November2002 (GElS), concludes that the environmental impacts of radiological accidents potentially resulting from decommissioning activities are of small significance and that this evaluation of their significance is applicable to all permanently shutdown units, including Humboldt Bay Power Plant (HBPP) Unit 3. Specifically, the GElS conclusion on the potential impacts of radiological accidents resulting from decommissioning activities states that, "with mitigation procedures in place, the impacts of radiological accidents are neither detectable nor destabilizing. Therefore, the staff makes the generic conclusion that the impacts of non-spent fuel-related radiological accidents are SMALL." For radiological assessments, impacts are of small significance if the Total Effective Dose Equivalent (TEDE) to a member of the public does not exceed the EPA 400 Manual protective action threshold of 1 rem, a small fraction of the limit established in 10 CFR 100.

Postulated accidents have been analyzed for HBPP Unit 3, independently from the NRC staffs evaluation, to address a site specific radiological issue that was not considered in the GElS. The accidents were selected in accordance with NUREG/CR-0672, "Technology, Safety and Cost of Decommissioning a Reference Boiling Water Reactor."

The following is a brief discussion of the HBPP site specific issue.

At the time that Unit 3 entered commercial service in 1963, the fuel utilized stainless steel clad. The stainless steel clad experienced gross faiJures during operation. These failures were severe enough that radioactive fuel was released from the clad and dispersed throughout numerous plant systems, contaminating these systems with alpha emitting radionuclides, i.e., uranium and transuranic isotopes. Some external plant surfaces have also been similarly contaminated.

HBPP completed the transition from stainless steel to zircaloy assemblies in 1969.

Over the many years since Unit 3 last operated in July 1976, beta and gamma emitting radionuclides have decayed, and alpha has become a more dominant factor in the potential dose contribution. Because alpha causes more severe biological damage when internal exposure occurs, the potential radiological dose consequences are likewise more severe.

This issue leads to a plant-specific potential environmental concern for Unit 3 decommissioning.

Sections 1.1 through 1.11 provide a brief description, key assumptions, and summary of the results for each analyzed decommissioning accident. Mitigating measures in the form of administrative controls will be in place (other equivalent controls are acceptable), where appropriate, to minimize the potential radiological environmental impacts of decommissioning activities and to maintain them within regulatory limits.

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All accident scenarios, contain surface contamination with the most limiting mixture of radionuclides (i.e., resulting in the highest total specific activity) as a component of the airborne release mixture. This limiting mixture is based on a sample of Spent Fuel Pool (SFP) crud per PG&E Calculation NX-322 (Reference 2.1 ). The mixture includes alpha emitters and other radioactive nuclides typical of BWR surface contamination. Some accident scenarios contain certain longer lived fission products, and neutron-activated plant metals and concrete as components of the release, depending on the decommissioning activity.

Detailed calculations, descriptions of the evaluations methodologies, and the mixture of radionuclides released for each of the decommissioning accidents in this appendix can be found in PG&E Calculations NX-323 through NX-333.

1.1 Dry Active Waste (DAW) Fire The following is a brief summary of the postulated DAW fire scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-323 (Reference 2.2).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences of a container (B-25 box) filled with DAW generated in alpha zones. Alpha zones are established at HBPP for the purpose of controlling contamination in areas having a high alpha to beta/gamma contamination ratio. Reference 2.1 identifies the limiting mixture of radionuclides used in this analysis and PG&E calculation NX-321 (Reference 2.3) identifies the maximum allowable radionuclide specific activities for qualifying the B-25 box for shipment as a Type A package.

Three different release scenarios have been considered in this analysis

• Filtered stack (elevated) release

• Un-filtered stack (elevated) release

• Un-filtered ground level release (fire in an open yard area)

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Assumptions The following key assumptions have been made:

The contents of the B-25 box are uniformly and totally consumed by the fire within 15-minutes of fire initiation.

Elevated releases, filtered or unfiltered, take place through the 50 foot plant stack.

A 99%> removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7o/o, which reduces the airborne release by a factor of 0.01.

The NUREG/CR-0672 combustible waste fire accident analysis fraction of 1.5E-04 of the radioactivity in the package is released by the fire.

Results Filtered Stack Release -

The TEDE to a member of the public for a filtered stack release is 3.97E-06 rem which is less than the EPA PAG limit of 1 rem.

Unfiltered Stack Release-The TEDE to a member of the public for an unfiltered stack release is 3.97E-04 rem, which is less than the EPA PAG limit of 1 rem. Therefore, the plant stack HEPA filters maintain the dose consequences ALARA (as evidenced by the filtered release results), but they are not required to keep the dose consequences within the EPA PAG limit for this postulated DAW fire scenario.

Unfiltered Ground Level Release -

The TEDE to a member of the public for an unfiltered ground level release is 4.01 E-03 rem, which is less than the EPA PAG limit of 1 rem.

Impact to Radiation Worker-If a radiation worker is present in the room for the entire 15 minute duration of the fire without respiratory protection, the exposure to the worker would be 665 DAC-hours, which equates to a TEDE of 1.66 rem. The EPA PAG limit does not apply to a radiation worker; however, 10CFR20 radiation worker limits would not be exceeded.

A-3 REV 10

1.2 Explosion of Liquid Propane Gas (LPG) Powered Equipment The following is a brief summary of the postulated LPG explosion scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies and the graphical depiction of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixture, see PG&E Calculation NX-324 (Reference 2.4 ).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences of an explosion caused by LPG mixing with air to produce a flammable mixture. The source of the LPG would be leakage from liquid propane powered equipment. The explosion results in destruction of plant stack HEPA filters and the adjacent ventilation ductwork. The explosive overpressure releases the accumulation of plant surface contamination and neutron-activated concrete dust from the HEPA filter and ventilation ductwork. One scenario occurs inside the plant with the airborne release discharging to atmosphere through the plant ventilation stack without the mitigating benefits of HEPA filtration. A second scenario assumes that the explosion damages the plant ventilation boundary and leads to an unfiltered ground level release.

Assumptions The explosion occurs when the stack HEPA filters and associated ductwork have the maximum radionuclide inventory.

Results Administrative Controls-The following administrative controls maintain the Total Effective Dose Equivalent (TEDE) to a member of the public less than the EPA 400 Manual protective action threshold of 1 rem. It is also acceptable to implement other controls that are judged to either be equivalent or that would preclude the accident scenarios from occurring.

• Use electric equipment in lieu of LPG-operated equipment (such as forklifts) to preclude the conditions in this accident (i.e., an explosion from LPG- powered equipment).

• Where concrete surface contamination is removed prior to LPG-powered equipment operations that could lead to a decommissioning accident through the stack, the TEDE to a member of the public is 1.35E-04 rem.

1.3 Vacuum Filter Bag Rupture The following is a brief summary of the postulated vacuum bag rupture scenarios, including a short description, key assumptions made, and results.

A-4 REV10

For all assumptions, as well as the detailed calculations and evaluation methodologies and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-325 (Reference 2.5).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences from the rupture of a vacuum filter bag filled with waste generated in Unit 3 alpha zones. Alpha zones are established at HBPP for the purpose of controlling contamination in areas having a high alpha to beta/gamma contamination ratio.

Sharp objects, such as metal shards, could rupture a vacuum filter bag during surface cleaning operations using a vacuum cleaner. Therefore, if a vacuum is used in alpha zones and the filter bag should rupture, there is a potential of generating significant airborne radioactivity. When the filter bag is ruptured, a conservative portion of the collected material becomes airborne depending on the particular scenario. For example, all of the collected material in the bag is assumed to become airborne during a building release because of the motive forces of the vacuum cleaner air flow.

Three different release scenarios have been considered in this analysis:

• Filtered stack (elevated) release

• Un-filtered stack (elevated) release

• Un-filtered ground level release (open yard area)

Assumptions The following key assumptions have been made:

Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99°/o removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7o/o, which reduces the airborne release by a factor of 0.01.

The vacuum bag contains the equivalent of 30,000 grams (66 pounds) of SFP crud (Reference 2.1 ).

For a vacuum bag rupture inside a ventilated building, vacuum cleaner air flow results in all 30,000 grams becoming airborne.

A-5 REV 10

Results Filtered Stack Release -

The TEDE to a member of the public for a filtered stack release is 4.57E-02 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

Administrative Controls -

The following administrative controls limit the alternate scenario consequences to within EPA 400 Protective Action Guide of 1 rem. It is also acceptable to implement other controls that are judged to either be equivalent or that would preclude the alternate accident scenarios from occurring.

Unfiltered Stack Release -

• Restrict the maximum vacuum filter bag contact dose rate to 2.02 rem/hr to maintain the maximum radiological impact to less than EPA 400 Protective Action Guide of 1 rem.

Unfiltered Ground Level Release -

• Restrict the maximum vacuum filter bag contact dose rate to 0.20 rem/hr to maintain the maximum radiological impact to less than EPA 400 Protective Action Guide of 1 rem.

1.4 Contamination Control Envelope Rupture The following is a brief summary of the postulated rupture scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-326 (Reference 2.6).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences of a contamination control envelope rupture.

If a contamination control envelope ruptures during oxyacetylene cutting, small globules of molten metal can.project up to 9 meters from the cutting operation. It is postulated that such molten particles penetrate the plastic sheet walls of the envelope and increase the leakage from the contamination control envelope from 10 percent to 50 percent. The result of the rupture is an airborne dispersion of dust consisting of Spent Fuel Pool crud and activated metals.

The scenario occurs inside the plant with airborne radioactivity discharging to atmosphere through the plant ventilation stack HEPA filters.

A-6 REV 10

An alternate scenario occurs inside the plant with the release to the atmosphere through the plant stack without HEPA filtration.

Assumptions The following key assumptions have been made:

Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99°/o removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7°/o, which reduces the airborne release by a factor of 0.01.

The increase in envelope leakage occurs during the removal of the reactor vessel when the highest specific activity material is being cut.

Since this task is remotely controlled, the large leak is ongoing for one hour of cutting before it is detected.

A scenario with the airborne radioactivity discharging directly to atmosphere at ground level is not credible because the airborne radioactivity is generated inside the plant ventilation boundary.

Results Filtered Stack Release -

The TEDE to a member of the public for a filtered stack release is 2.52E-3 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

Unfiltered Stack Release-The TEDE to a member of the public for an ufiltered stack release is 0.252 rem which is less than the EPA 400 Protective Action Guide of 1-rem.

1.5 Oxyacetylene Explosion The following is a brief summary of the postulated explosion scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-327 (Reference 2. 7).

Description Oxyacetylene cutting torches may be used for removing and segmenting various steel components during dismantlement.

A-7 REV10

Oxyacetylene explosions can occur from such causes as flow reversals, nozzle obstructions, and flashbacks (a flare going back up the gas hose).

3 The explosion occurs while operating oxyacetylene cutting tools within a 1000 m contamination control envelope. The result of the explosion is an airborne dispersion of dust consisting of neutron-activated reactor vessel steel and surface contamination. The accident also releases additional carbon steel from six destroyed HEPA filters.

The scenario occurs inside the plant with the airborne radioactivity discharging to atmosphere through the plant ventilation stack HEPA filters.

An alternate scenario occurs inside the reactor building with the radioactive dust from the explosion going through the plant ventilation system without any HEPA filtration and discharging to atmosphere through the plant stack.

A third scenario considers the unfiltered release at ground level.

Assumptions Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99°/o removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7o/o, which reduces the airborne release by a factor of 0.01.

Results Filtered Stack Release The TEDE to a member of the public for a filtered stack release is 7.29E-03 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

Unfiltered Stack Release The TEDE to a member of the public for an unfiltered stack release is 0. 729 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

Administrative Controls-The following administrative controls preclude the third ground level and unfiltered scenario from occurring. It is also acceptable to implement other controls that are judged to either be equivalent or that would preclude the alternate accident scenario from occurring.

• Use arc cutting tools rather than oxyacetylene to avoid the accident scenario or

• Assure that all oxyacetylene usage will be conducted in such a manner that no damage could occur to the plant stack or pressure boundary due to an explosion of the oxyacetylene.

A-8 REV 10

1.6 Filter Damage from a Blasting Surge The following is a brief summary of the postulated blasting surge scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-328 (Reference 2.8).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences of HEPA filter damage due to a blasting surge. The historic background for this accident is the filter damage that occurred during the Elk River reactor decommissioning due to concrete blasting activities. The radionuclide inventories are neutron-activated concrete and concrete surface contamination. As concrete demolition occurs, the filter initially accumulates dust from the contaminated surface layer; subsequent drilling produces dust from activated concrete.

Any accumulated neutron-activated concrete dust and surface contamination are released from the HEPA filter by the explosive overpressure.

The scenario occurs inside the plant and involves damage to a local HEPA filter installed as an RP control for the concrete blasting work. Airborne radioactivity from the damaged filter discharges to the atmosphere through the plant ventilation stack HEPA filters. An alternate scenario occurs inside the plant with the release to the atmosphere through the plant stack without HEPA filtration.

Assumptions A scenario with the airborne radioactivity discharging directly to atmosphere at ground level is not credible based on the source of airborne radioactivity being inside the plant ventilation boundary.

Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99°/o removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7%, which reduces the airborne release by a factor of 0.01.

Demolition blasting of the concrete biological shield walls is in progress. The explosion occurs near the end of the demolition so that the filter has a maximum radionuclide inventory.

14,000 grams (31 pounds) of neutron activated concrete dust are released from the damage filter. A much smaller mass of surface contamination is also released.

A-9 REV 10

Results Filtered Stack Release The TEDE to a member of the public for a filtered stack release is 1.39E-03 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

Unfiltered Stack Release The TEDE to a member of the public for an unfiltered stack release is 0.139 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

1.7 Detonation of Unused Explosives The following is a brief summary of the postulated detonation scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-329 (Reference 2.9).

Description The purpose of this analysis is to evaluate the offsite radiological dose consequences of the detonation of unused explosives. The explosive charges are onsite for the purpose of removing the biological shield. An accidental explosion occurs inside the Refueling Building with contamination shields and water spray off, causing concrete destruction and releasing a dust cloud of surface contamination.

The resultant airborne radioactivity is discharged to atmosphere through the plant ventilation stack HEPA filters. An alternate scenario occurs inside the plant with the release to the atmosphere through the plant stack without HEPA filtration. A third scenario discharging directly to atmosphere at ground level is considered.

Assumptions Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99% removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99. 7o/o, which reduces the airborne release by a factor of 0.01.

The radionuclide release consists only of surface contamination, since the surface is the most likely material to become airborne in this scenario. The surface contamination is much greater in terms of activity than the structural concrete below, making this is an acceptable simplification.

50 grams of concrete surface contamination become airborne.

A-10 REV10

Results The TEDE to a member of the public for a filtered stack release is 7.59E-05 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

The TEDE to a member of the public for an unfiltered stack release is 7.59E-03 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

The TEDE to a member of the public for a release assumed to occur at ground level with no filtering is 7.68E-02 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

1.8 Minor Transportation Accident The following is a brief summary of the postulated minor transportation scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-330 (Reference 2.1 0).

Description Accidents involving trucks carrying radioactive waste from a decommissioning site may result in the release of radioactive material. The purpose of this analysis is to evaluate the offsite radiological consequences of a minor transportation accident involving the release of radioactive material generated at the site. The scenario involves an airborne release from a fire and occurs to a truck shipment of waste containing the bounding mixture of radionuclides (Reference 2.1 ). Class A limits as defined in 10 CFR 61.55 and Utah Administrative Code R313-15-1 008 were used to determine the bounding specific activity for each resident isotope.

Assumptions As a result of a traffic accident, a Type A 55-gallon container catches fire and a fraction of 5E-04 (.0005) of the bounding activity derived in Reference 2.3 becomes airborne.

A member of the public is located 100 meters downwind from the fire.

Results 55-gallon container fire The TEDE to a member of the public is 4.54E-03 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

A-11 REV 10

1.9 Severe Transportation Accident The following is a brief summary of the postulated severe transportation scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-331 (Reference 2.11 ).

Description Transportation accidents involving trucks carrying radioactive waste from a decommissioning site may result in the atmospheric release of radioactive material. The purpose of this analysis is to evaluate the offsite radiological dose consequences of a severe transportation accident involving radioactive waste generated in alpha zones at the site. The analysis considers an intermodal container catching fire.

Reference 2.1 identifies the limiting mixture of radionuclides used in this analysis and Reference 2.3 identifies the maximum allowable radionuclide specific activities for shipment as Type A packages.

Assumptions As a result of the traffic accident, an intermodal container catches fire and a fraction of 5E-04

(.0005) of their activity becomes airborne.

A member of the public is located 100 m downwind from the accident.

Results The TEDE of 3.04E-1 rem to a member of the public is less than the EPA 400 Protective Action Guide of 1 rem.

1.10 HEPA Filter Fire The following is a brief summary of the postulated HEPA filter fire scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-332 (Reference 2.12).

Description A fire in a portable ventilation unit HEPA filter connected to an alpha zone contamination control envelope has the potential of generating significant airborne radioactivity. The purpose of this analysis is to evaluate the offsite radiological dose consequences of a fire in a HEPA filter being used in this configuration.

Three different release scenarios have been considered in this analysis.

• Filtered stack (elevated) release

• Un-filtered stack (elevated) release

• Un-filtered ground level release (fire in an open yard area)

A-12 REV10

Alpha zones are established for the purpose of controlling contamination in areas having a high alpha to beta/gamma contamination level. In this analysis a portable ventilation unit is connected to a contamination control envelope and is processing material being generated within unit alpha zones. A typical portable filtered ventilation enclosure unit consists of a large blower coupled with a HEPA filter preceded by a glass-fiber roughing filter, all mounted on a wheeled cart. A flexible duct couples the cart unit to the contamination control envelope; the envelope surrounds the work area and confines the materials being generated. Roughing filters are installed at both the inlet and the outlet of the ventilation enclosure unit.

Assumptions The HEPA filter is loaded with radioactive material with the limiting radionuclide mixture of Reference 2.1.

Elevated releases, filtered or unfiltered, take place through the plant stack. ,

The entire filter is uniformly consumed by the fire during a 15-minute period.

Results Administrative Control Based on current radiological practices at HBPP, the filter is changed out at a contact dose rate of 1 mrem/hr.

Filtered Stack Release The TEDE is 8.16E-06 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

Unfiltered Stack Release The TEDE is 8.16E-04 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

Unfiltered Ground Level Release The TEDE is 8.25E-03 rem which is less than the EPA 400 Protective Action Guide of 1 rem.

A-13 REV 10

1.11 Loss of HEPA Filtration The following is a brief summary of the postulated filtration loss scenarios, including a short description, key assumptions made, and results. For all assumptions, as well as the detailed calculations and evaluation methodologies, and the graphical depictions of the relationship between TEDE and the total curies released for the relevant radioactive nuclide mixtures, see PG&E Calculation NX-333 (Reference 2.13).

Description Loss of portable ventilation unit HEPA filter capability while connected to an alpha zone contamination control envelope has the potential of generating significant airborne radioactivity. The purpose of this analysis is to evaluate the offsite radiological dose consequences of a loss of HEPA filtration capability while the filter is being used in this configuration.

Three different release scenarios have been considered in this analysis.

• Filtered stack (elevated) release

• Un-filtered stack (elevated) release

• Un-filtered ground level release Alpha zones are established for the purpose of controlling contamination in areas having a high alpha to beta/gamma contamination level. In this analysis a portable ventilation unit is connected to a contamination control envelope and is processing material being generated within unit alpha zones. A typical portable filtered ventilation enclosure unit consists of a large blower coupled with a HEPA filter preceded by a glass-fiber roughing filter, all mounted on a wheeled cart. A flexible duct couples the cart unit to the contamination control envelope; the envelope surrounds the work area and confines the materials being generated. Roughing filters are installed at both the inlet and the outlet of the ventilation enclosure unit.

Assumptions Elevated releases, filtered or unfiltered, take place through the plant stack.

A 99°/o removal efficiency is conservatively assumed for plant stack HEPA filtration rather than the design value of 99.7o/o, which reduces the airborne release by a factor of 0.01.

The unfiltered release would continue for a 15 minute period.

30,000 g (66 pounds) of the radionuclide mixture identified in Reference 2.1 is released in each scenario.

Results Filtered Stack Release The TEDE is 3.43E-02 rem, which is less than the EPA 400 Protective Action Guide of 1 rem.

A-14 REV10

Administrative Controls The following administrative controls limit the alternate scenario consequences to within regulatory limits or preclude their occurrence. It is also acceptable to implement other controls that are judged to either be equivalent or that would preclude the alternate accident scenario from occurring.

Unfiltered Stack Release

• Incorporate HEPA filter redundancy into the ventilation unit to preclude an unfiltered stack release on the loss of a single HEPA filter or

• Limit the contamination control envelope Am-241 concentration to less than 1.06E-05 1-JCi/ml.

Unfiltered Ground Level Release-

• Incorporate HEPA filter redundancy into the ventilation unit to preclude a ground level release on the loss of a single HEPA filter or

• Limit the contamination control envelope Am-241 concentration to less than 1.05E-06 1-JCi/ml.

2.0 References 2.1 PG&E Calculation NX-322, "Decommissioning Accident Analysis-B, HBPP SFP Crud Sample Activity" 2.2 PG&E Calculation NX-323, "HBPP Dry Active Waste Fire Analysis" 2.3 PG&E Calculation NX-321, "Decommissioning Accident Analysis- A, HBPP Crud Waste Class A Limit" 2.4 PG&E Calculation NX-324, "Explosion of LPG Leaked from a Front-End Loader Accident Analysis" 2.5 PG&E Calculation NX-325, "Vacuum Filter-Bag Rupture Accident Analysis" 2.6 PG&E Calculation NX-326, "Contamination Control Envelope Rupture Accident Analysis"

2. 7 PG&E Calculation NX-327, "Oxyacetylene Explosion Accident Analysis" 2.8 PG&E Calculation NX-328, "Filter Damage from a Blasting Surge Accident Analysis" 2.9 PG&E Calculation NX-329, "Detonation of Unused Explosives Accident Analysis" 2.1 0 PG&E Calculation NX-330, "HBPP Minor Transportation Accident Analysis" 2.11 PG&E Calculation NX-331, "HBPP Severe Transportation Accident Analysis" 2.12 PG&E Calculation NX-332, "HBPP HEPA Filter Fire Accident Analysis" 2.13 PG&E Calculation NX-333, "Loss of HEPA Filtration Accident Analysis" A-15 REV10

Enclosure 2 PG&E Letter HBL-14-003 Humboldt Bay Quality Assurance Plan, Revision 31

Nuclear Power Ge.ner.ation NUMBER L-4 VOLUME 4 Humboldt Bay REVISfON 31 EFFECDATE :6-/-13

• Power Plant PAGE 1 ort*

TITLE APPR~~~

HUMBOLDT BAY QUALITY ASSURANCE PLAN (liBQAP) 6, lir~n s~~ :U4??13 DIRECTORJPLANT NAGER I DATE HBNUCLEAR (Procedure ClassificaOon- Quality Related)

J.0 DESCRIPTION This pJan was originally submitted to the NRC as Attachment4 ofthe application for the "SAFSTOR" Hcense amendment "to allow possession out not operation ofthe reactor. It was subsequently modified to include those elements required to implement Independent Spent Fuel Storag~ Installation (HB ISFSl) activities und.er 10 CFR 72 Subpart G (previously the applicable portions ofthe Diablo Canyon Quality Assurance Prograrn). This document d¢scribes the Quality Assurance Plan to be applied to Humboldt Bay Power Plant Unit J during SAFSTOR and decornm)ssioning, and includes HB ISFSI

~ctivities. The HBQA.P satisfies the requirements of 10 *cFR 50 A,ppendix.B, 10 CFR 71 Subpart H, and 10 CFR 72 Subpart G. .

2.0 ATtACHMENTS 2.1 Humboldt Ba~ Quality Assurance-Plan (HBQAP)~ RevisionlL 3.0 RESPONSIBLE ORGANIZATION Director~ Qu~lity Verification

L-4 Attachment 2. J Rcv.31 Page 1 of 52 HUMBOLDT BAY QUALITY ASSURANCE PLAN (llBQAP)

TABLE OF CONTENTS Secti()n Title Page INTRODUCTION *2 1.0 ORGANlZATJON 3 2.0 QUALITY ASSURANCE PROGRAM 9 3 ..0 DESIGN CONTROL 11 4.0 PROCUREMENT DOCUMENT CONTROL 13 5.0 INSTRUCTIONS, .PROCEDURES,. AND DRA WfNGS 14 6.0 . DOCUMENT CONTROL 15 7.0 CONTROL OF PURCHASED MATEIUAL~ EQUlPMENT, AND SERVICES 16 8.0 IDENTIFlCATJON AND CONTROL OF MATERlALS, PARTS, AND COMPONENTS 18 9.0 SPECiAL PROCESSES" 19 10.0 INSPECTJON 20 11.0 TEST CONTROL 22 12.0 CONTROL OF MEASURING AND TEST EQUlPMENT 23 13.0 HANDLING; STORAGE, AND SHIPPING 24 14.0 INSPECTION, TEST,AND OPERAT1NG STATUS 25 15.0 *coNTROL.OFNONCONFORMING CONDITIONS 26 16.0 CORRECTIVE ACTlON 27 17.0 QUALITY ASSURANCE RECORDS 28 18.0 AUDiTS 31 APPENDIX A* APPLICABLE STRUCTURES, SYSTEMS, AND COMPONENTS 35 B ADMINrStRA TIVE CONTROLS 37 c COMMITMENTS 45 D ORGA1',HZATION CHART 52

L-4 Attachment 2.1 Rev. 31 P~gc2 of 52 HUMBOLDT BAY QUALITY ASSURANCE PLAN (HBQAP)

INTRODUCTION Pacific Gas ~nd Electric (PG&E) has established and is implementing a comprehensive Quality Assurance Program for Humboldt Bay .Power*Piant Unit 3 (HBPP Unit 3) and Humboldt Bay Independent Spent Fuel Storage I.nstaUation (l-IB ISFSJ) to assure conformance with established regulatory requirements set forth by the Nuclear Regulatory Commission (NRC). This Quality Assurance Program is identified as the Humboldt Bay Quality Assurance Plan (HBQAP). This

. HBQAP has been cleveloped to assure quality assurance controls are effectively implemented fqr HBPP Unit 3 SAFSTOR Operations and decommissioning activities, Radioactive Materials Trarispoi1 Packages anc;l the operation of HB ISFSI.

• Th.e HBQAP is a comprehensive statement of the-quality assurance controls that govem the ovetall

• decommissioning ofHBPP Unit 3 and operation and maintenance ofthe*HB JSFSI. The HBQAP describes the organizational structure, f-unctional responsibilities, lev*els of authority, line of communication, and interfaces.

• The requirements and commitrrients contained in the HBQAP are mandatmy and *must be implemented, enforced, and adhered to by all individuals and organizations performi~g activities affecting quality. Workers are encouraged to actively participate in the continued improvementand implementation of the HBQAP. Any necessary changes should be promptly communicated and*

implemented~

The HBQAP satisfies the "requirement~ of 10 CFR ~0 Appendix B; 10 CFR 71 Subpart Hand 10 CFR72Subpart G. The HBQAP applies* to all activities associated with structures, systems, and c.or:nponents (SSCs) wh~ch are important to $afety defined within 10 CFR 12. The HBQAP *also applies to transportation packages licensed by the NRC: under 10 CFR 71 and limited decommissioni~g activities (10 CFR 50,Appendi:x B). The implementation of the requirements of theHBQAP is perfmmed in a graqeq approaqb comrraensurate withan itejils or activitie~ importance to safety. This graded approach is appllc~ble to important to safety SSCs and is responsive to NRC Regl)latory Guide 7.1 0. The applicability ofportions of the HBQAP to other items and activities is detennined on a case-by-~ase basis as defined In Appendix B.

The HBQAP Is impie111ented through the use of approved procedures (i.e. policies, procedures, manuals, instructions; or other documents) :that provide written guidanpe for the control ofimportant t() safety items al)d .activities. The B.BQAJ> ~lso: prcwid¢s for development ofdocurnentafion to demonstrate objective evidence of compliance with the stated requirements.

• L-4 Attachment 2.1 Rev.31 Page3 of 52 1.0 ORGANIZATION PG&E~s efforts are to assure the quality and safety of the SAFSTOR Qperations and decommissioning of the HBPP Unit 3, the packaging arid tr~nsportation ofr~dioactive materials and the operation and maintenance of the JIB ISFSI utilizing the HBQAP. The program is organized in a structured manner with clearly defined levels ofauthority, assignments of respo_nsibHH:y and lines of communication. As~ignment of the responsibility for an item or activity includes responsibility for its, quality. The Quality Verification (QV) organization providesindependent oversight of qua-lity activities and maintains lines of communication with Senior Management.

PG&E has assumed full re~p 0 nsibility to its employees, stockholders, the general publ_ic and affected governmental regulatory agencies for the establishment and execution ofthe* HBQAP prescribed herein, quality-r.e1a.ted _program directives and administrativ~ procedures. The work of

.executing selected portions of the HBQAP may be delegated to organizations external to PG&E; however, in ~ll.s~ch ir~stan*ces, PG&E r.etains overall responsibility.

Specific responsibilities pertaining to ql1ality as*~mrance matters ate assigned by the HBQAP *and its Implementing procedures and instructions to various individuals throughout:PG&E. In-each instance, the ~ssignmeh.t of.a responsibilitY to. an individual includes with ita commens~rate delegation of sufficient authority that the person can, in fact, fulfill that responsibility. Unle~s othetwise specifically prohibited, it is und-erstood that the functions, tasks and activities necessary to carry out a responsibility n;ay be delegated to a11d performed by other qualified in-dividuals.

All delegations of functions, tasks, activities and authority shall be documented.

The reporting relationships associated with the individual responsibilities discussed in this section are *sh.own in part in App~n-Qix D. This.QA Phm uses generic titles and identiftes:functions:and responsibilities for those titles used in* this document. Differences (if any)- to actual tities used in the or~anization .are traceable to. the :QA .Plan_ titles. by the use of a.dministr.ative procedures.

The BOARD OF DIRECTORS - PG&E CORPORATION is responsible for all facets of PG&E*s utility busfness.

The..PRESIDENT(genedc title) is acco.untab1e to the Board of Directors and establishes the corporate p*otic.ies, goals, and .objectives related to ali .ofPO&E's ac.tivitiesand*operati"ons. the President is responsibJe for the phmnif!g, ~:listdbu.tron and-dev¢1opm:ent of.:all the. Comp~ny's energy. resources arid nuclear power generation .. These function*s:inClude;engineering,

• qonsttu.cti.o.n~ HB~P Unit3 dec.ornmiss:icin.in_g? and HB 'ISFSI opetutions.

the SENIOR VlCE. *PRES.iDENT-- ENERGY SUPPLY reports to: the Pres.idenl Pacific Gas and E.lectd9 CQnipany and. ovf;!tSe~s .an ofPac.mc GaS. an<! Electric Company~s..energy proct.Jtement and power generation assets, inc1uding nuclear~ fossil and hydroelectric, cogeneration and r¢newable: sources; w:ith a. focus oh phmni.ng .for a-nd delivering reliab.l6., ~ffordable and clean electricity and *gas for distdl?utioq_t(): ~a~jfi'c (J@.S ~nd :EJ.e~tri.c Comp~ny-,custorner~..

The SENIOR VICE Pltf:SJPENT,,. Cl-JJEf NUCLEAKOFfiCER (g¢n.er;i~- title) ,(CNO) is.

responsibJefor overallHB:ISFSI :and HBPPUnit 3 decommissionil'l:g*safety and for:taki:ng.

measutes.needed to .ensure a¢c.eptab.le performance ofthe 1-iB ISFSl staff in de*signhig1 fabricati*ng, *c*onstructing,: t~s(in g;- ~p~rat.ing, m.o*difying, deco.mmis.s.io_n in'g,,and providing techn'(c~[ iupport: tO *th,e iffi *ISFSi and~ dudng Jffipp Unit J decommissioning.

lr4 Attachment 2.1 R~v. 31 Page 4 of 52 The CNO, or designee, as specified in administrative procedures, approves and signs official company correspondence to the U.S*. NuClear Regulatory Commission (NRC) or its represen.tatives. The CNO approves revisions to the QA Program as described herein that constitute a reduction in a commitment made to the NRC.

The VICE PRESIDENT- ENGINEERING (generic title) is responsible for providing, upon request: (1) technical investigations, tests, analyses, examinatiQns. and calibration services in support of the HBPP Unit 3 and the HB ISFSI; (2) developing, evaluating, qualifying, testing, and improving welding, brazing, and. heat-treating procedu~s required by the company; ~md (3) prpviding evaluation suppor;t*of these procedures.

The VICE PRESIDENT -SUPPORT SERVICES (generic title) is responsible for providing document services support for the HBPP Unit 3 and the HB lSFSI. These services inc-Jude indexing, preparing, and duplicating microfiche for the drawing con.trol system; storing the master microfiche and drawings that cannot be microfilmed; and scanning and indexing drawings when requested. Also responsible for administering, coordinating, p.hmning, and operation of warehousing and procurement ofmaterla1s in support ofHBPP Unit 3 and HB ISFSI operations ai1d const.ruction, as well as for contract services.

The EMPLOYEE CONCERN$* PROGRAM SUPERVISOR(generic title) provides an acceptabl~ *alternative path, independent of the line management process, for company and contractor employees to report safety concerns and concerns dealing with harassment, intimidation, retaliation or di~crimination without fear of retaHation for having engaged in protected activity.

The DIRECTOR- QUALITY VERIFICATION (generic -title)(~V Director) is responsible for assuring that th.e HBQAP ~nd adminjstnitive pto,c:edures are,effecti:Vely implemented and complied with by all involved. organizations~ both internal and external to PG&E. The Oii'¢ctor QV has been given the Qrganizational freed,om and delegated the requisite authority. to investigate any area or aspect of PG&E's HBPP Unit 3 decommissioning and HB ISFSI operations as necessary to identify arid define problems associated with *establishment or execution of the HBQAP; Thedireetor*also has been delegated the authority to *initiat~,recommend or provide solutions for such problems to whatever management level is necessary, and to verify that effective corrective*action is taken .i.n a timely manner, Thi_s deleg(,ltion includes the authorlty to assess, -review, inspect,. audit; and monitor the conduct of quaHty;.related actlv1ties performed by or* for *PG&E~s HBPP Unit 3 and HB ISFSI to assure compliance with the HBQAP and ~ther

• regulatory requir~.m~nts*.

• The QV Directorhas:acces_~ to the President, the CNO,-the HB:Director, and &ppropriat¢ m~flagers for any significant quality.:related problem or deficiency~ .Th~ Q.V Dir¢ctQr*is

• authorized to. prescribe a uniform com,pany--wide method ofpetforming' an activity affecting.

q1,1_ality by Sponsofing Ot requirjng the. i~$lJCii}Ce of procedt;r¢s wb_el} SUC.h stan(:Jard.ization is considered desirable or essential to the effectiveness ofthe HBQAP. *Such uniform methods are co.ntained in adtnh1isttative-programs and pro.cedure~, and-compliance with their requitemenlsby all PO&E and contract personnel is mandatory. *

• L-4 Attachment 2.1 Rev. 31 Page 5 of 52 The QV Director will not be responsible for any activities unrelated to responsibilities described in the HBQAP that would prevent the required attention to QA matters. The responsibility of the implementation of the Hl3QAP will take precedence over the other non-Q~ duties.

The QV Director is responsible to regularly assess and report on the status, adequacy and effectiveness ofPG&E's HBQAP to the CNO and other affected PG&E management and oversight comrnitt~s, and for providing recommendations*on.solutions to quality problems.

The QV Director is responsible to identifY, prepare and submit for approval such changes to the HBQAP prescribed. herein as are necessary t9 maintain the HBQAP up to date and in ..

conformance with current regulatory requirements and PG&E commitments to the NRC. The QV DJrec.tor is responsible for the review of all regulatory submittals as they pertain to the HBQAP and hjs/her concurrence is required prior to submittal.

The QV Director .is responsib.le for performing monitoring, as-sessments, independent QC inspections, reviews and audits to verifY that the activities covered l>y the HBQAP are performed correctly, with minimal human error. This includes activities associated with HB ISFSI *design, fabrication, constru.ction, testing, operation, modification, decommissioning and related :activities; HBPP Unit 3 decommissioning activities governed by the HBQAP; and supplier quality. The QV Director has the authority and responsibility to stop work should. there b.e .a serious breach of any part of the HBQAP or of technical or regulatory requirements. wherein public health or safety could.be involved, The bire9tor assures timely and effective corrective actions through audits, regular assessments and qu.a.lity assessment status repot1s.

The QV Director is responsible for verifying conformance to established requirements (except designs) is accomplished by individuals or groups within QV who do not have direct responsibility for perf<)rming the work being verified or by individuals or groups tr~ined and

.qualified in QA concepts-and practices and independent of the organization responsible for*

performing* the ~sk. The persons and prganizations performing QA ao9 qua1ity .contr<>l functions have direct access to management levels that assure the*abilityto: (a} identify quality problems; (b) initiate, recommend, Qr provide solutions through designated channels; and {c) verify implementation of solutions. They are sufficiently free from direct pressures for cost and schedule and have the responsibility to stop unsatisfactory work and control further processing, delivery, *c>r instaJiation of nonconforming material. The organizational positions with stop work authority are identified in th.e rmplementing procedures. .Reporting to the QV Director are the q~ality assurance? quality "verification,_ supplier quality? project quaiity and independent quality control inspeoth;>'n functi,oos.-

The Dl.RECTOR -HUM_BQLDT BAY (generic title) (H.B Director)js re~poosibl~ for* the conduct*

of all activities related to the.HBPP Unit 3 decomm"issioning:and the HB isFSI. This includes responsibility for operation~ maintenance, engineering, radiation protection,_ training; and security.

The~ bir~~tod§ re$p.onsible for the Iodependent Safe!)! Reviewer (I.SR) proce:ss. The HB birect¢ris responsible .for the development of, and is authorized to. approve and dire.ct the implementation of; those. ptograms, proc~dures, and insttucti()os r¢quired for ibe :AJ3lSFSl and HBPP Unit 3 within the. limits estab1ished by this HBQAP,the HBPP.Unit.3 and HB ISFSI.

Technical Spedfications, and adtninistraUve guideHn*es *established in the fiB ISFSI Fi"nal Safety

-Analysi~ Report (FSAR) ~nd 'HBPP Unit 3 Defueled Safety AnaJy$is Report (D.SJ;\R). Design authority for theJIBPP t.Jn'it 3. and HB ISFSI has also been delegated to:him. *The t:t!.l P.irector shall be accountalfle for adherence to the operating lim.l.ts and requite*ments contained in the lSFSI and HB_PP Unit 1 Tecbnic~l ~ife9.ifications. The fill Dir~tor sh.~ll be r¢.sponsi,ble for-the operational command: function. * **

L-4 Attachment 2.1 Rev. 31 Page 6 of 52 The HB Director shall delegate these responsibilities to other specific members of the p.lant staff during his/her absence.

The ENGINEERING MANAGER (generic title) is responsible for technical aspects of the engineering and des.ign ofHBPP Unit 3 and HB 1SFS~ SSCs established for monitoring system peiformance and trends; for performance of modifications to the Humbo1dt Bay HB ISFSI; for configuration control and design bases defense and management; for quality classification of SSCs; and for the specification of technical and quality requirements for the purchase of services, materials and equipment.

The ISFSI.MANAGER(generic titl~) is responsible for the operation and maintenance.ofthe HB ISFSJ. .

The RADIATION PROTECTION MANAGER (generic litle) is responsible for implementing the radiation protection program at HBPP Unit 3 for the protection ofthe workers and members of the public.

THE NUCLEAR SAFETY OVERSIGHT COMMITIEE (NSOC) 'implements the Independen~

Review described *in Appendix B.

INDEPENDENT SAFETY REVlEWER(S) (ISR) performs independent safety reviews as defined in Appendix B.

Each organization that: supports the HB JSFSI and the HBPP Unit 3 documents and maintains current a written desoriptibn of its internal organiZation; This documentation describ~s the or b~siness unit department's structure, levels.of authority, lines of communication, and assignments of responsibility. Stich documentation takes the form of organization charts supported by written job descriptions or other narrative material in sufficient*detail that the d~ties and authoritY of e~ch individual whose work affects quality .is clear.. Interface~ between organiz~tions are de~icribed in ~drn.inistrative procedures or other <)ocoooents ~oritrolled in accordance with the appropriate requirements.

The individuals assigned to the positions having a particular responsibility in administrative procedures (as described above) are the only individuals who are iu1thorized to perform these activities. J-Iowever~*drcumstanc~s may a~i~e where iti~ considered eitb¢r neces~ary or desirable to have such activities, or some portion ofthem,.actually performed by someone*else. In such ca$es, the ~ss.igning organi~tion teta.ins respons.ibiUty and shall. verify that t.he proc*edures and io,stfuctionsto be f~llowed jn performing the work are adequa~e for controlling*the wo.rk*and meet.

• applicable requirements. In such circumstances, the detailed procedures and* :instructions to be foJ lqW.ed in performing the W9tk are *reviewed andjtpprovecl by the person 'a$signed r~sponsibHity for the work .prior to the commencement of work. The purpose of su.ch* review and approvai is to verify that such procedures-and "instructions reflect .an acceptable method of performing the work and ar~ *in. compliance with fh~ requ'irements c:>f the Hl3QAP. All in*stances* in:~ which authority is to be delegated: or support services are to be provided,_ are documente~ .

. I

L-4 Attachment 2.1 Rev. 31 Page 7 of 52 Defined quality suppliers to HBPP Unit 3 and the HB ISFSI are required to conform to the HBQAP or to their own program approved by PG&E. Supplie~'!QA Programs are required to comply with ~he applicable portions of 10 CFR 50, Appendix B, 10 CFR 7_1; Subpart Hand 10 CFR 72, Subpart G, and the applicable regulatory documents and industry *Standards identified in Appendix C. The quality program is defined in the contract or similar procurement document.

Suppliers to PG&E are required to document their internal organizational arrangements to the extent necessary for PG&E to-assure the supplier is capable of effectively managing, directing, and executing th.e requirements of the procurement documents. The authority and responsibility of persons and organizations who perform activities that might affect the quality of the procured ite.ms or services shall be clyarly established. The Suppliers' organizational stntcture, levels of authority, and functional assignments of responsibility shall be such that:

(I) The QA function of formally verifying conformance to the teqhnical and quality requirements of the procurement documents is accomplished by qualified personnel who

.are independent of those who performed or directly supervised the work.*

(2) Personne.l who performQA functions have sufficientauthority and organizational freedom to identify quaiity problem~; to initiate, recommend, or provide solutions; to verify implem*entation ofthost! solutions; and to control further processing of the items or services until a proper disposition has occ1,1rred.

Facility Staff Qualifications Except as specified in other portions of the HBQAP, each member of the facility staffshal.l meet or exceed the fo11owing minimum qualifications:

Fot Hl3 Unit 3, the minimum qualifications of ANSI N 18.1 (197:1) for comparable

. _positions-are applicable.

For the HB ISFSI, the minimum qualifications of ANSI I ANS 3.) (1.978) for comparable positions are applicable. In addition, HB ISFSl personnel shall meet the requirements of the HB ISFSI Training Program.

The following requirements and clarifications apply to the p\)shions listed below, applicable to both BB Unit3 and the HB ISFSI; The Radiatio*n Ptotecticm M.anager. shall meet o~ exceed th¢ tequirements of Regulatory Guide 1.8, R~vision **2, April, 19.87.

The QV Ditectot shall meet. th.efollowing qualification requirements:

management experienc.e through assignments to responslb1e.positions; k.now1edge.of QA regt.llations, policies; practices, and standards; and .experience, working in *QA *or related activity in reactor design~ construction, or operation or in a similar highly technological industry. At the time of-assignment to the active position, the QY Director shall have six years of experience in ii:npl~menting quality assurance, pre,fer~bly at an: op_erating nucle~r plan. tor operati.ons supervisory experience.

L-4 Attachment 2.1 Rev. 31 Page 8 ofS~

At least on~ year of these six years of experience shall be nuclear power plant experience in the overall implementation of~ Quality Assurance.Program (QAP).

A minimum of one year of this six-year experience requirement shall be related tec~nical or academic training. A maximum of four years of this six-year

• experience requirement may be fulfilled by related technical or academic training.

For the QV Director, the one year of qualifying nuclear power plant experience in the overall implementation oftheQuality Assurance program can be obtained out.side the Quality Assurance organizations.

L-4 Attachment 2.1 Rev. 31 Page 9 of 52 2.0 QUALITY ASSURANCE PROGRAM 2.1 PROGRAM APPLICABILITY The HBQAP. requirements, as a minimum,. apply to those HBPP Unit 3, Radioactive Materials Transport Packages and HB lSFSI SSCs defined in Append*x A and satisfies the requirements of 10 CFR 50 Appendix B, I0 CFR 71 Subpart H and 10 CFR 72 .Subpart G. *The HBQAP ~pplies to the design, fabrication, construction, testing, operation, maintenance, modification and decommissioning ofthe HBPP'Unit3 and HB ISFSI structures, systems, and components (SSCs) defined in (\ppendix A.

The applicabJe QA criteria are applied to an exten.t that is commensurate with the .importance to*

safety.

The effectiveness of the implementation of the HBQAP shall be assured through Quality Related programs and administrative. procedures.

Additional requirements for management and administrative controls are described in Appendix B .of the HBQAP to ensure the safe decommissioning ofthe Hi3PP Unit 3 and the operation of the

• HB ISFSI throughout the life of licensed activities. These controls will provide confidence that HBPP Unit 3 and HB ISFSI SSCs will perform satisfactorily in service, including activiti~s that determine that physical characteristics and quality of materil!ls or components adhere to predetermined requirements.

2.2 . PROGRAM CONTROL The status and adequacy of this HBQAP shall be regularly monitored and it shall be revised as necessary to improve its effectiveness or to reflect changing conditions.

The HBQAP documents, including any changes, supplements, or appendices.are*issued and maintained* as controlled doc.uments. Changes to the HBQAP reqt,~irements shall be made in accordanc*e with 10 CFR 50.54. Changes to the HBQAP that do not .reduce commitments shall be included in the periodic updates requlred by 10 CFR 50.71. Proposed changes*to this HBQAP that tedu*ce commitments are re"Yi¢wed and concurred with in. WriJing by the QV Dlrec.tor and a.re*

approved by the CNO~ or designee, prior to being submitted to and approved by the NRC in accordance wlth l 0 CPR 50.54 prior to issue for use; lmpl¢ment~tiol) of th~: HBQAP is ~ccomplished throqgh separa_tely iss~ed procedures, ins.tructions, and drawings. Each vice president, director, and manager is .responsible for the establishment and implementation of detailed pwcedures and instructions pre*scribing the activWes for which they are responsible. Such documents ar~ derived from the requirements and reflectthe resp.onsib.llities specified in the HBQAP. Ac.tivities affectingqu.alicy are accomplished in accordanGe w:ith these instrliGti<;>lls,. procedures, and drawingsAll personnel. are in$tructed that compJiance-with those requirements and the re*quirements ofthe:HBQAP is* mandatory.

• Questions ofdi.sputes involving interpretations ofHBQAP requirements or of the commitments*

a net reqvirements~ upon which it is based, ~re referred to the QV I)irecto:r for .r~sOlution, Questions or _di.sputes involving the responsibilities defined in the HBQAP are referred to the CNO;

L-4 Attachment 2.1

• Rev. 31 Page 10 of 52 Questions or disputes involving other quality matters are resolved by referring the matter in a timely manner to successively higher levels of management until, if necessary, the matter reaches that level which has direct authority over all contesting parties.

• The QV Director or designated representative, provides regular reports to the CNO, responsible comp.any management, and NSOC on the effectiveness ofthe HBQAP a.s it relates to HBPP Unit 3 decommiss.ioning and HB lSFSJ op¢r~tions. Such tepor(s are based on the results of audits, reviews, *inspections, tests, and other observations of activities as prescribed.by the HBQAP:

Tl1e QV Director shall repot;t annually to the CNO on. the effectiveness of the QA Program and results ofthe.Audit Program. The repott shall include an evaluation of compliance with current regulatory requirements and commitments to the NRC.

2.3 INDOCTRINATION AND TRAINING Personnel- involved in implementing the activities within the scope of this HBQAP shaiJ be responsible* for the quality of their work.. Tbese personnel shaH re.ceive:

• Indoctrination in the requ iremeilts of this HBQAP;

• Indoctrin*ation in their organi:zation's irnpleme.il~ing pro.cedures; arid

• Tt~ining a':ld qualified in tasks requiring special skills or knowJedge.

Indoctrination, training, qualificati.on, and re-qualific.ation (when applicable) shall be p~~s~ribed and performed in accordance with written procedures, in accordance with applicable codes, standards (lnd regulatory requirements, which specify the.t1'1anage(lleQt responsibilities; training areas; frequency of training; method of qualification and requalific.ation; and ~ocumentati.on teql.lirements.

L-4 Attachment 2.1 Rev. 31*

Page 11 of 52 3.0 DESIGN CONTROL Design Control applies to those SSCs that are designated as important tb safety described in Appendix A> which suppo~ the HB ISFSI and HBPP Unit 3. Design activities shall be performed in an orderly~ planned, and controlled manner to achieve applicable HB ISFSI and HBPP Unit 3 design related activities that best serves the needs ofPG&E.and its citstomers without posing an und~e risk to the he*aJth and safety of the pubJic. Design activities shall be.controUed to assure that design, technical, and quality requirements are.correctly translated into design documents and that changes to design and design documents are properly controlled. Design control procedures shall address respoqsibilities for all phases o(design including: * *

( 1) Responsibilities

• (2)

• Interface control (3) Design input (4) Design performance

. (5) Design verification

• (6) bes.ign change Systematic methods shall be established and documented for communicating n~eded design informatipn across the external and internal design interfaces, including .changest() the design information> as work progresses. The interfaces between the supporting PG&E engineering organization and otheH+IBPP Unit 3 and HB ISFSI organizations, either intemal or extern:aJ to PG&E, performing applicable work affecting quality of design shall be identified and documented. This identification shall in.clude those organizations providi.ng criteria, designs>

specifications; technical direction, andtechnical infonnation and shall be in sufficient detail to cover each applicable SSC and the corresponding design actiVity. -

Provisions for design input shall define the technical objectives for SSCs being designed t>r analyzed. For the SSC being designed, or for the design services being provided (for exampJe, design verification); design input requirements shall be determined, documented, reviewed, approved, .and controlled. Required design an*aJyses (such as stress, them1al, hydra.LJlic and accident analysis; materiai compatibility; maintenance and repair and ALARA *considerations) shall be performed in a planned, controlled, and correct manner. PG&E procedures shall identify the review'~nd approval responsibilities for design analyses.

The preparation and ~antral ofdesign docutnents (such as specifications, drawipgs; reports, and.

instaJiation procedures) $h~ll be performed in a manner to assure design jnpUt$*3re correctly translated into design documents (for example, a documented check to verify the .dimensio_n~l accuracy and completeness of design drawings- and specificatit>ns). PG&Esball provide for reviewing, confinnirig, or substa_ntiating the design to assure that the design meets the specified design inputs. *Design verification shall be* performed. by competent individuals or groups other than thos~ who performed the originaJ desi~n> bt

Jt who may be frQm the. same dep~rtment

• Individuals performing the verification shall.not:

(1) Have immediate supervisory responsibility for the individual performing the .design. In exceptional circumstances, the designefs immediate supervisor can:per{orm the ver,jfication provided:

L-4 Attachment 2.1 Rev. 31 Page 12 of 52 (a) The supervisor is the only technically qualified individual (b) The need is individually documented and app1;oved in advance by the supervisor's management (c) Quality assurance audits cover fr~qtHmcy and effectiveness of ~tse of supervisors as design verifiers to guard aga_inst abuse (2) _Ha:ve specifi~d_ a singular design ~pproach (3) Have ruled-out certain design considerations (4) Ha~e established the design inputs for the particular d~sign aspect being verified The results of the design verification efforts shall be documented with the identification of.the verifier cl~arly provided. Design verification methods may incl.ude~ but-not be Hmited to, the followi_ng: design reviews, use of alternate *calculations~ and qualification testing. The design .

verification shaH he identified and documented. The design verification shalJ be completed* prior to relying upon the compon~nt system or structure*to perform its function. Pr()cedures shall assure that- verified computer codes are certified for use *and that their applicability is specified ..

Proposed c_hanges-or modifications to the applic_abl.e SSCs defined in Appendix, A shall be designed

-by a qualified Individual or organization~ and reviewed by a qualified individual/group other than the *individual/group who prepared the change or modification, but who may be from the same organization. These reviews shall incfude a determination as to whether additional cross-discipline reviews a:re necessary. lfdeerned necessary, they sba11 be p~rfonued by review pers.onnel of the appropriate discipline($). These reviews shall also dete1111ine whether an evaluation per 10 CPR 50 59 or i 0 CFR 72.48 is .necessary. If necessary, one shall be prepared and provided to¥-an Independent Safety Reviewer for. review prior to approva I.

Each change or modification to tbe de_signated SSCs i.n Appendix A sha)] be approved by the H8 Director, or.designee, as specified in adm-inistrative procedures, prior to implementation.

Proced_ures for implementing design changes.; including field ch~mges> shall assure that the impact of-the chan.ge is carefully considered, requited a.ction~ documented, and information c;Qil~_erning the change transmitted to all affected persons and *organizations. These. changes shall be

$ttbjected to* design c9*ntrol measures commensorate: with those applied tq the ori~htal*design.

Design changes shall ber¢viewed and approved by the same organization or -group that was responsihle-for the .original design.

Documeritcontrol measures shall be established for desig;fl documents that reflect the.

co.mmitments of the applicable desigp. ba~is. document the:se- d¢sJg*n docunHmts: shall include specificati'oos, .calcQ]Jltions,compu.ter program~, _system, de~cript_ions, tb~ a.pplic_a~le de$ign l:>asis do.cument (when used as a design document),, drawings .(including flow diagrams, piping* and ins_trumel)t diagrams, c.qntrollogic di;xg:ram$, e.l~ctricfil $Ingle lin¢ ()iagn).ms? 'St_r'llcfural dtaWi_J)gs for majorfaciiities, site arrangements), and equipment locations.

Nonc()nforming .a..c.tJy'jties $l]ch .as prQced,ure vi.o_l~ti.~ms, deviations, or *err~rs_ a.ri.d ,d~t:icie~cies .in approved design *docume~ts, "including design methods (such as co0.1,puter codes), shall he conkolled as des~r.ibe-Q 'intSe¢tions 15.0 and 16;0,

L-4 Attachment 2.1 Rev. 31 Page 13 of 52 4.0 PROCUREMENT DOCUMENT CONTROL Procurement Docuntent. Control applies to. those SSCs that a~e designated as importantto ~afety described *in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP Unit 3. The procurement documents shall include those requirements necessary to assure thanhe items and servi9es to be provided will be of the desired quality.

The procurement documents shall also inc Jude* provisions for the fol_lo~ing, as appropriate:

(1) Basic Technical Requirements -These includ¢ drawings, specifipation~, codes, and industrial standards with applicable *revision data; test and inspection requirements~ and special instructions and requJremeiits, such as for designing, fabricating, c.Ieaning, erecting, packaging,_ handling; shipping, and, if ~pplicable, extended storage in tbe field.

(2) Quality Assurance Requirements- These include the requirements for the supplier ~o h*ave-an acceptable Quality Assurance Program(QAP); provisions for access to the supplier's facilities ~nd records for source i.nspection and audit when the n.e*ed for such inspection and audit has been .determined; and provisions for extending applicable.QAP

• and other requirements of procorement documents. to su.bcontractors and suppliers, including *pa&E's access to facilities and records.

(3) Documentation Requirements- These shall include recprds to be prepared, maintained, submitted oi* made avai"lable for review* and instructions on record retention and di~position.

The procedures that implement pi*ocur~menl docurnent control shall describe the organization.al responsibilities for procurement planning; preparation, review, appt;oval and control of pt*ocurerrient documents; supplier setectio"n~ .bid ev'aluation*s; .and "review and .evaluatio*n ofthe suppfier's: QAP prior to initiation of"activities affected by the program.

Procedures shall be established to review the adequacy of technical ~nd quality assurance requil.'ements stated in procurement documen_ts; *determine that. requirements are*correctiy stated, inspectab~e*, .and controllable; assttre ad~qu*~*~ acq~pt~nce ~nd rejeption C,riteri~; al)d pr~vid~ for the preparation,_ review, and approval of procurementdocuments in accordance with HBQAP requirements. The.*.review*and docume.nted:concu.rtenc.e of.the adequacy of technical,- and quaJity assurance-require,uents st~ted in procur~m~nt documents sh~ll be performed by independent in personnel trained and .qualified applicable QA practices and concepts.

Changes to procurement documents shall be subject to the same control as th~orig.inal do(;ument.

L-4 Attachment 2.1 Rev. 31 Page 14 of 52 5.0 INSTRUCTION

S. PROCEDURE

S, AND DRAWINGS.

Controls for Instructions, Procedures, and Drawings apply to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP Unit3. Activities applicable to the SSCs defined in Appendix A shall be prescribed by and accomplished in accordance with documented instructions, procedures, and drawings. AdditiQnal guidance for other administrative controls is provided within Appendix B.

The HB Director is responsible for the establishment and implementati.on of instructions, procedures, or drawings prescribing activities~required by the itBQAP. Standard guidelines for the format*, content, and review and approval processes shall be established and set forth in a procedure or instruction. The method of performing these activities shall be prescribed in documented instructions, procedures; or drawings of a type appropriate to the circumstances.

This may include'shop drawings, process specifications, job descriptions, planning sheets, travelers, QA manuals, *checklists, or any other written or pictorial fonn provided th~t the activity is described in sufficient detail such that competent personnel could .be expected to satisfactorily perform the work functions Without direct supervision.

Within the constraints, limitations, or other co~ditions as may be imposed by{the specific license requirements or commitments, procedures prescribing a preplan ned method of conducting activities required by the HBQAP shall be established in accordance with the applicable

• regulations, codes, standards, and specifications.

• In addition, the above procedures and programs, which are required to support important to safety SSCs and associated activities shall be established and controlled as described in Appendix B.

Written p-rocedures shall be established, implemented, and maintained covering the activities referenced in the HB ISFSI and. HBPP Unit3 Technical Sp~cHications .. Each of these procedures,

~nd changes thereto, and all proposed tests or experiments that affect nuc.lear safety sh*au be reviewed and approved prior to implementation in accordance with the review and approval requirements in Appendix-B. Each of these procedures shall also be r¢viewed periodically as set forth in administrative procedures. A responsible organi~tion shall be assigned for each~of these procedures. The responsible organization shall assign reviews of proposed procedures, programs, and changes to qll~li[led personil_el of the appropriat~ discipline(s).

• The procedure rev.iew and approval requirert)entsdefined in AppendixB apply w~en ~pproving applicable HBPP Unit 3 and liB ISFSI programs and procedures or when maling changes .to those designated HBPP Unit 3 ancl HB ISFSI programs and *procedures. They also apply when approving or changing corporat~ procedures and procedures used by support organizations if they could have an immediate effect on HBPP Unit 3 or ISFSI operations or the operational status ofHBPP Unil3 or .lSFSI SSCs that are important to safety. They do not apply to editotial or typographicaJ changes.

L-4 Attachment 2.1 Rev. 31 Page 15 of 52 6.0 DOCUMENT CONTROL Document Control applies to those SSCs and associated activities that are designated as important to safety described in Appendi.x A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP ~nit 3 and the administrative controls described in Appendix B.

Documents and changes to documents that prescribe or verify activities affecting quality shall be controlled in a manner that precludes the use of inappropriate or outdated documents. As a minimum, controlled documents include: design documents, including* documents related to computer codes; procurement documents; instructions and procedures for such activities as fabrication,. constrUction, modificatiqn, installatlc:m, test, opera.tion; maintenance, and in~pection; as-built documents; quality assurance and quality control manuals and quality-affecting procedures, HB ISFSI FSAR updates, HBPP Unit 3 DSAR updates, and Significant Conditions Adverse to Quality/nonconformance reports.

The organization responsible for establishing instructions, procedures, drawings, or other documents prescribing activities affecting quality is also responsible to develop and implement systematic methods for the control o( sus;h documents in accordance with the requirements herein. ln those instances where such documents directly involve organizational interfaces, that organization with *ultimate responsibility for the issuance of the documents is responsible for establishing the methods for thejr controL Procedures and instructions shall assure that documents, including changt;s, are prepared; reviewed by a qualified individual other than ~he person who generated the document; approved for release by authorized personnel; distributed to the location where the activity is performed

pdor to commencing work; and lJsed in performing the activity, Proced1,1res and instru.ctions shall

.require the development ofas-built drawings and the removal or appropriate identification of obsolete or superseded docu*ments. . - .

Procedmes and instructions.thatdefine methods for implementing the BBQAP requirements shall be reviewed ~nd concurred with by quality verification (QV), for compliance and alignment with the Program. Revisions to these documents shall also be reviewed and concurred with by QV if they propose a change to the BBQAP as it is d~scribed in a commitment to a regulatory agency.

The. controls shall identifY those responsible for prep~ring, reviewing, .approving, arid issuing qocurnents to be used, They sh~ll also defi_ne the coordination and control ofinterfac_ing documents and shall require the establishment ofcurrent and updated distribution iists.

A document co~trol.system shall be established to identify the curr~nt: revision of in$trus;tions, procedures, specifications, drawings,. and procurement docum~nts. Master lists, when*utilized as an elemeoJ of the qocurnent control system, ,shall be upd~ted and distributed to predetermined re~ponsib1e personnel.

• L-4 Attachment 2.1 Rev. 31 Page 16*of52 7.0 CONTROL OF PURCliASED MATERIAL, EQUIPMENT, AND .SERVICES Control of Purchased Material, Equipment, and Services applies to those. SSCs and associated activities that are designated as important to safety described in Appendix A,_ which support the HB ISFSl, Radioactive Materials Transport Packages, and the I-ffiPP Unit 3. Supplie-r activities in providing purchased material, equipmen.t, and services shall be monJtQred ~_s planned an.d necessary to assure such items and services meet procurement document requirements.

Procedures shall descr"ibe .each organization's re$ponsibil ities for tbe control.ofpurchased m~terial, equipment, and seryices, including the interfaces between ail affected.organizations.

All materials, equipment, and services shall meet the specified technical and quality requirements. Verification that a supplier can meet the specified technical and quality requirements sha1i be by one or a combination of the following:

(1) Evaluation of the supplier's history*

(2) Evaluation of current supplier quality records (3) Evaluat~on of the *supplier's facilities, personnel, and implementation of~ Quality Assurance *Program (QAP)

Such evaluations shall be documented. A Supplier?s QAP that has been found by quality verification (QV), to ~atisfy specified quality requirem_ents shaH be listed on the PO&.E Qual_ified Suppliers List, which is controlJed by QV.

Suppliers ofcommercial grade calibration services may be qualified based on _their accreditation by a nattona'lly~recognized a:ccrediting.body, as. an alternative to qua:Hfi.catlon by supplier' audit~

commerqial*grade survey, or in-process surv~illance.

A d.ocument~d review of the suppliers' ac~reditation by.the*purchasermay be used as the qualification method, as described in PG.&E commitmentstoNRC Regulatory Gul~es 1.123 and

1. 1.44, which are docu~rnented in Appendix, C; Thi~ review shaH _include, ~ta. mi'nimtitn, ~.11 oftbe following:

( l) The a~cteditation is to ANST/I.SO/IEC 1704$

(Z) 1'he ~ccred.iting body is either the Nation_al Vohmtai)' Laboratory A.cqred.itatjo.o Pr()gram (NVLAP) or an accrediting body recognize.d ~y NVLAP. thr()ugh a:M.\ltQal Recognition Agreement (MRA). * *

(3) The published scope ofaccreditation forthe calibration laboratory cove,rsthe needed

measurement parameters, ranges, .and uncertainties.

A quality verHI*catlon plan shaft *be established and docttmented that-applies to *each procurement a:rtd id~ntifles the m.anner by which Pd&E. intends*(with appropriate QV orga.nizaHon or iQ,v.olvem~nt): to a$SUre the. quality of the material, eql,l_ipm~nt,. servi~~ ~~ d.efin.ed itXthe procurement documents and to accept:those hems or services from the supplier.

The .quality. verification plan shall identify Inspection, audh, and/or -surv.eiliance.activitles to be performed includhtg the .characteristics or processes to .be. witnessed, inspected, ;or verified; the rnethc)d ofsnrveill~~WY; and *~he extent ofdQ~um~Qtation. r~q~!in~d.

L-4 Attachment 2.1 Rev. 31 Page 17 of52 The timing and sequence of the activities shall be planned to identifY any system or product deficiencies before subsequent activities may preclude their disclosure.

The plan shali also.be based on consideration of:

(1) Importance to the HB ISFSI safety

• (2) Complexity .of inspectable characteristics (3) Uniqueness.ofthe item or service Supplier performance and compliance with pi*ocurement documents may be monitored by either source verification, t*eceiving inspection, or a combination ofthe two. Source verification

. activities may consist of inspections, audits, surveillance, or a combination thereof and are condllcted at the supplier's facility. When source verification activities are specified in the quality verification plan, the timing and sequence of these activities are to be delineated.

Receiving inspection .c;tctivities, as described in the .quality verification plan, sha II be coordinated with source verjficatio.o activities performed prior to shipments.. lf sampling is performed~ it shall be ih accordance with proc~dures and/or recognized standards. Receipt inspection shall include a review which verifies that supplier quality records required by procurement documents are acceptable and that items are properly identified and traceable to appropriate documentation.

Re.cords of quality verifi.cation ac.tivities shalt be traceabl_e to the materials, equipment; or service.$

to which they apply. Documentation *ofacceptance in accordance with the procurement quality verification plan shaH be ayaHable at the site prior to installatio"I1 or acceptance for use.

Documentary evidence that procurement document requirements have been met shall clearly refl~ct eac.h requirement. Supplier's Certificates of Confothianc¢ are _pe't'iodically *evaluated by au.dits and independent -inspectioos or tests to assure they are valid and the results docl_lmented.

When spare or replacement parts are procured, suppi iet selection 11nd *qu.al ity verification activities shall be. planned and implemented' to verifY compliance with requltem:ents meeting or exceedingthos.e of the odgin.at

L-4 Attachment 2.1 Rev. 31 Page 18 of 52 8.0 IDENTIFICATION AND*CONTROL OF MATERJALS, PARTS, AND COMPONENTS The Ide.ntification and Control of Materials, Parts and ComponeQts'applyto those SSCs and associated activities that ~e designated.as important to safety describelln Appendix A; which support the HB ISFSI, Radioa.ctiv¢ MaterialsTransport Packages, and the HBP:P Unit 3.

Materials, parts, and components shall be identified and controlled in a manner to preclude the use ofincorrect or defective items.

All materials, parts, and components, including partially fabri~ated subassemblies, batc}J.es,' lots, and corisurnables, shal I be idetltified in a manner that each can be related to its .applicable drawing, specification, or other teehnical.documentation at any stage from initial receipt through fabrication, installation, repair, or modification. Controls and implementing procedures shall ensure that Of'!ly correct and accepted items are w~ed during all stages and ~es*cribe the

• responsibilities of the involved organizations.

Physical identification of items shall be used whenever possible and practicaL Controls may, however, be through physical separation, procedt~J'e, or other appropr.iate means. Identification may be either on the item or on records traceable to the item.

Identification marking, where employed, shall he clear, unambiguous, *and indelible and its application. shall not impair the function of the identified item or any other* item. When an item is

.subdivided, th~ identifying marking sha:tl be transferred to each resulting part .. Markings shall not be rendered il.legible by treatrmmt, proc¢ss, assembly, install adon, or coat.ing unless other means ofidentification and determining a~ceptapility are provided. .

Veri[l(;}ttion activities; such as iospec.tion~ shall be perfonned to ensJ.Ir~ that the provi~iqns of this policy and related implementing procedures are followed for items pr'ior to release for fabrication, assembly, shipping, i.ostallat.ion.* and use.

When required by code; standard, or specification, traceability of materials, parts, or components

• tO' specHlc *inspection or test recor~s shall be ptovid~d for and verified ...

L-4 Attachment 2.1 Rev. 31 Page 19 of52 9.0 SPECIAL PROCESSES l

Special Processes applies to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP Unit 3. Special processes shall be controlled and performed by quaJified personnel using qualified procedures or instruction.s in accordance with appJicable codes, standards, specifications, criteria, or other special requirements.

A special process is an activity in which the quality of the result is highly dependent upon either pr~cess variables or the skill. and performance of the p,erson doing the work, and the specified quality is difficult to verify by inspection and test after the process is completed.

Special processes include, but are not Iimited to:

(1) i Welding (2) Heat treating (3) Nondestructive examination (4) Chemical cleaning (5) Others as specified in design and procurement documents (examples are certain protective coating applications and concrete batch plant operati.ons, Which are controlled by specifications on a case-by-case basi's)

The implementing instructions shall conta*in .the criteria for assuring proper process contro1 and shaH be qualified and controlled to assure compliance with applicable codes, standards, QA procedures, and .design specifications. Sub~tantiating records of qualifications and controls shall be maintained. *

• L-4 Attachment 2.1 Rev. 31 Page20 of52 10.0 INSPECTION Inspection activities apply to those SSCs and associated activities *that are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP Unit 3. A comprehensive program of inspection of items and activities affecting quality shall be conducted to verify conformance with established requirements.

Procedures shall describe the organizational responsibilities necessary to carry out the inspectio~

program.

The objective ofthe; inspection program shall.be to verify the quality ofthe items and activities and conformance to the applicable documented instructions, procedures; and drawings for*

accomplishing activities affecting quality. The inspection program, including information

• relative to individual inspections to be performed, shall be developed based on a review of the design drawings, specifications, and other controlled documents which prescribe ite~s and activities affecting quality. Inspections shall be performed utilizing appropriate inspection procedures and instructions together with the necessary drawings, specifications, and other controlled documents. The inspections shall be documented and evaluated.

Inspection procedures, instructions, or checklists shall provide for the following: identification of qharacterjstics and activities to be inspected; a description ofthe method Qf inspection; identification ofthe individuals or groups responsible for performing the inspection *operation; acceptance and rejection criteria; identification of required procedures, dra\vings~ and specifications and revisions; recording the name of the inspector or data recorder and the results of the inspection operation; and specifying necessary measuring and test equipment including accu~acy requirements. The inspection program shall include~ but not be limited to those inspections requjred by applicable codes, standards, specifications, and the HB ISFSI Technical*

Specifications.

The itJspect.on program shall require inspections o.fHB ISFSJ and HBPP Unit 3 modifications, repairs, and replacements to those appJicabJe SSCs designated in Appendix A to be in accordance with e~isting design requirements. *These inspections will be planned anq implemented in accordance with approved procedures*.

• The inspection program shall*require inspection and/or test of items for each work operation where such is necessary to assure quality. Ifinspection of processed Items is impossible or disadvantageous, indirect control by monitoring of process shall be required~ Both inspection and process monitoring ~hall be required whet) control is inadequate without both. Both inspection a:nd process control shall be performed when required by applicablecode,.standard, or specification.

Mandatory quality. control inspection hold points shall be identified *in the* inspection program.

When. required,. the :specific hold poin.ts shall: b~ iijdic~teq in the drawings~ procedure.s., or instructions that prescribe the work activity. Work shall not proceed beyond such hold points wjthout the docunien:ted consentofQuality Verification. When the inspection *program *perm[ts or requtres a sample of a la:rge groliP of items that are ~men able to statistical analysis, the s.ampling procedures to be u§ed -shall be based o.n recognized stand~rd practices,

• Inspections to verify the*q~ality of work shall be performed by qualified ind.ividuals otber*than those who performed or directly supervised the.activity being inspected. During the inspection, su~h perso.ns shaH not. report directly to the immediate sup.ervisors *wh.o, ar~ re$ponsible for ~he work bei"ng *inspected.

L-4

. Attachment 2.1 Rev. 31 Page 21 of 52 .

Personnel performing inspections shall be qualified in accordance with applicable regulations, codes, standards, and specifications.

Inspection r:ecords shall contain the following where applicable: a description of the type of observation, the date and results of the inspection, information related to conditio11s adverse to quality, inspector or data recorder identification, evidence as to the acceptability of the results, a:nd action taken to resolve any discrepancies noted.

L-4 Attachment 2.1 Rev.31

. Page 22 of 52 11.0 TEST CONTROL .

Test Control applies to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI and the HBPP Unit 3. A program of testing shall be conducted as necessacy to demonstrate that SSCs will perfonn satisfactorily in service. This program shaJJ ensure that the necessary testing is identified and performed at the appropriate time in accordance with written*test procedures that inc-orporate or reference the requirements and acceptance limits contained in the applicable design documents.

Thy program shall cover all ~equired tests, including t~sts prior to- installation, preoperational tests, and operational tests.

The procedures that implement testing shall provide for meeting appropriate prerequisites for the test (for example, environmental conditions, specification ofinstrumentation, and completeness of tested item)t sufficient instruction for the performance of the test, specification of any witn.ess or hold points, acceptance and rejection criteria and limits, and-the documentation *of the test.

The procedures shaH proVide for evaluatio_n and documentation -of the test results and data and their acceptability as detennined. by a qualified person* or group.

Test records shall contain the foJJowing where applicable: a description of the type of observation, the date and results of the test, information related to conditions adverse to qu111ity, inspector or data recorder identification, evidence as to the acceptability of the results, and action taken to resolve any discrepancies noted.

L-4 Attachment 2.1 Rev. 31 Page23 of52

. 12.0 CONTROL OF MEASURING AND TEST EQUlPMENT The Control ofMeasuring and Test Equipment applies to those SSCs and associated activities that are designated as-important to safety described in Appendix A, which*supp01t the HB ISFSI, Radioa-ctive Materials Tr~nsport *Packages, and the HBPP Unit3. Organizational responsibilities shaiJ be delif)eated for establishing, implementing, and assuring the effectiveness of the calibration program for measuring and test equipment (M&TE). This program shall include the generation~ review, and documented concurrence of calibration procedures; the calibration of measuring and test equipment~ and the m~intenance and use of calibration standards.

M&TE, including reference*standards, used to determine the acceptability of items or activities shall be strictly maintained within prescribed accuracy limits.

M&TE, including refer~nce standards> shall be of suitable range, type, and *accuracy to verify conformance with requirements.

Procedu-res for control of M&TE shall provide for the identification (labeling, codes, or alternate documented control system), recall> and calibration (including documented pre-calibration checks) of the M&TE_- The-calibration procedures shaiJ delineate any necessary environmental controls, limits, or compensations in excess of those which may be inherent to the general program.

The calibrations shall utilize documented valid relationships to nationaiJy recognized standards or accepted vaJues of natural physical constants. Where national standards do not exist, the basis for the calibration shall be documented. Calibration of M&TE shall-be against standards that have an accuracy of at least four times the req\}ired accuracy of the equipm_ent being c.alibrated or, when this is not practical, have an accuracy that assures the equipment being ca11brated will be within required tolerance and tharthe basis of acceptance i~~documented and authorized by responsible management of the PG&E organization performing that activity.

Calibrating standards have greater accuracy than standards being calibrated. Calibrating standards with the same accuracy may be used if it can be shown to be -adequate for the requirements and the basis ()facceptance _is docum~nted and authoti_z¢d by responsible management.

The calibration interval~, whether calendar- or usage:..based, shall be predetermined and documented. lndication of expiration) if feasible, will be displayed on or with the M&TE.

Signi:ficarit e.nvironn:te.ntal or usage restrictions wi11 be ind_ic~ted OJJ or with the equipment or be-factored into the documented system used to control t_he issuance ofthe M&TE. Special cali~ratio.n shall be required whenever t_he accuracy of the equipment is suspect.

R~cord$ shall be maintained to show that estabHshed schedules and procedures for the calibration of the M&TE have been foJiowed. M&TE shal I be identified and traceable to the calibration test

• data. Records *of the usage oftlw M&TE shall be m~intained to facUitate..corre9tiVe acti9n in the:

eventof the discovery of a deficiency concerning the ca:Hbration or use of M&TEt so that measures may be taken_ qncJ doc~men~ed tp determimtthe validity ofptevit>us inspecti"on~

performed and oqhe acceptability ofitems inspected or tested *since the previous caJibration of the deficient M&TE~ * *

• L-4 Attachment 2.1 Rev. 31 Page24 of 52 13.0 HANDLING, STORAGE, AND SHIPPING Handling, Storage and Shipping requirement~ apply to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and th~ HBPl' Unit 3. Material and equipment shall be handled, stored, and shipped in .accordance with design and procurement requirements in a manner that will prevent damage; deterioration, or loss. .

Special coverings, equipment, and protective environments shall be specified and provided where necessary fqr the protection ofparticl,llar- items froin damage or deterioration. . When suqh special protective features are required~ their existence shall be verified and monitored as* necessary to assure they continue to serve their: intended function. - .

Special handling tools and equipment shall be provided where necessary to ensure items can be handled *Safely and without damage. Special handling tools and eq*uipment shall be controlled and maintained in a m~nrier such that they will be ready and fit to serve their intended function when ne*eded. Such contl'ol shall include periodic inspection and testing to verify that special

• handling tools and equipment have b.een properly maintained.

SpeCial attention shall be given to marking and labellng items during packaging, shipment, and storage, Such additional marking or labeling *shall be provided as is necessary to ensure that

. items can be properly maintained and preserved. This shall include indication ofthe presence of special environ*ments orth~ ne_ed for special control. Provisions shall be*described for the storage of chemicals, reagents (iQcluding control*of shelf life), lubricants, and other consumable materials.

Special handling, preservation;.stonige, cleaning, packaging, and shipping: requirements are established and accomplished by suitably trained. individuals in accordance with predetermined work and inspection instructions.

J..,-4 Attachment 2.1 Rev.31 Page 25 of 52 14.0 INSPECTION, TEST, AND OPERATING STATUS Inspection, Test and Operating Status requirements applies apply to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI and the HBPP Unit 3. The inspection, test, and/or operating status of material, equipment, and operating sys.tems shall be readily apparent and verifiable. ,

The procedures used to indicate status shall provide means for assuring-that required lnspections and tests are perfonned in the prescribed sequence; acceptability is indicated; and nonconforming

,items are clearly identifiep throughout fabrication, installation, test, maintena.nce, repairs, and modification to prevent inadvertent use or operation. Items accepted and released ~re identified to indicate their inspection status priorto forwarding them to a contro)ledstorage area or releasing them for installation or further work. Deviations from the prescribeq sequence shall be subject to the same level of control as the generation of the original sequence to prevent the bypassing or omission of a required test o*r inspection.

Identification of status may be by such means as, but not limited to, tags, stamps, markings, 1abels, or travelers. ln some instances, records*traceable to the item may be used. The procedures implementing controJ of inspection, test, and operating status shall clearly delineate authority for the application, change, or removal of a status identifier.

L-4 Attachment 2.1 Rev. 31 Page26 of 52 15.0 CONTROL OF NONCONFORMING CONDITIONS The Control ofNon~onfonning Conditions app*Jies to those SSCs and associate.d activities that* are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materia1s Transport Packages, and the HBPP.Unit 3. Items and activities that do not confonn to requirements shall be controlled in a mann~r that will prevent their inadvertent use.or installation. Technical decisions as to the disposition of each nonconforming condition shall be made by personnel with assigned authority in the relevant discipHnes. The control, review, and disposition of nonconforming conditions ~hall be accomplished and documented in accordance with approved wt:itten procedures and instr!Jctions .

.Nonconforming-conditions shall be documented and affected organizations notified -of such conditions. Further processing ofthe nonconfol11)ing conditions and other items affected by them shall be controlled in a manner to prevent their inadvertent use or installation pending a decision on their disposition.

• The responsibHity and authority for the disposition of nonconforming conditions shall be established and set forth in the applicable procedures and instructions for their control. The rework or repair ofnonc(>nfonning. items and .th..¢ dispo.sition of operational" nonconforming conditions shall l;>e accomplished in accordance with written procedures and instructions.

Dispositions involving design changes shall be approved by the organil.ation With the authority for design.

• The acceptability of rework or repair of material$, parts, components, systems, or stnJcrures shaiJ be verified by re~inspecting and retesting the item as originally inspected and tested, or by a method that is at least equ~l to the origina.l. inspection or ~esting method. Reworked anq repaired items shall be* re-inspected in accordance with applicable procedures and instructions. The acceptability of nonconforming items that have been dispositioned "repair" :or "accept-as-is" shall be documented. *such ,documentation shall include a description of the change> waiver, or deviation that *has been accepted in order to record. the change and, if applicable, deriote the as-built condition.

• Corrective action for conditions adverse-t.o qu.aUty*shaJt be processed in accordance with Section 16.0.

. In c~es where req\)ire<i documentary evid<mce that items have passed required. inspectionS' and tests is not available, the associated materials or equipment shall be considered .nonconforming.

Until suitable:d~curnentary evidence is availabl~jo s.hQWthat the mate.dal or equiprn¢nti"si.n be conformance, affected systems shall be considered to inoperah.ie and reliance shall not be placed .on such systems to fulfill the it intended 'safety functi.ons.

NonconformiJ1g conditions that requ.lre reporting to the, NRC shall be reviewed *by the QV organization. Such ,feview sh.all include the results ofany i.nvesti.gations made and the

. reconu1te.ndations f¢sulting from such investigati()ns to pr~clude or reduce the probability of recurrence* of the ev.ent or circumstimce. ** * **

• L-4 Attachment 2.1 Rev. 31 Page27 or52 16.0 CORRECTIVE ACTION Corrective Action applies to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB lSFSI, Radioactive Materials Transport Packages, and the HBPP Unit 3. Each individual condition adverse to quality subject to the HBQAP shalJbe identified, COiltrolled, and evaluated, and a dispos'ition shall be determined for the remedial action and corrective action as soon as* practicable. These activities shaH be performed consistent with Section 15.0, Control ofNonconfonning Conditions. These me~sures shall be established in written procedures and utilized for documenting, reviewing, .and dispositioning of these qu!!_lity problems.

Systematic review and evaluation of all conditions adverse to quality shall be conducted and documented. Conditions adverse to quality sh.all include, but not be. limited to: engineering, design, and drafting errors; equipment failures and maifunctions; abnormal occurrences; deficiencjes; devia~ions; and defective material, equ1pm~nt, and services.

The revieW and evaluation shall include identification of quality trends, repetitive occurrences, and significant conditions adverse to quality. The quality trends and other significant review findings shall be analyzed and appropriate corrective action determined. Findings and actual or recqmmended corrective action shall be reported to management by the responsible organization for review and assessment.

• Significant conditions adverse to quality shall be investigated to the extent necessary to assess the root c~uses and to determine the corrective action required to preven~ recurrence ofJhe.same*or similar conditions. The corrective action required for significant conditions adverse to quality shall. be accomplished in a timely Immner. Significant condition.s adverse to qu.ality, the cause of the condition, and the corrective action taken shall be documented and reported to management.

Signi:fica~t conditions adverse to quality shaH be repqrted to the Quality Verification organization.

Completion ofcortectiv~ actions for significant conditions adverse to .quality shall*be reviewed and verified by personnel having no direct responsibility for either the disposition or the corrective

• action taken~ Follow-up reviews shall be conducted to verify that the corrective action was properly implemented, performed in a timely manner, and th~t it was effective in correcting the

- identified condition.

Significant conditions adverse to quality shall be evaluated for reportability to the NRC in accordance with 10 CFR21, IOCFR 50.72, 10 CFR 5033, 10CFR50.9, 10 CFR 72.74, and 10

.CFR 72,75~ the HalSFSI T~chn.ica;l Specifications and other applicable ~egul~t-ons and .shall be reported as required ..

L-4 Attachment 2.1 Rev. 31 Page 28 of 52 17.0 QUALITY ASSURANCE RECORDS The requirements for Quality Assurance Records apply to those SSCs and associated activities that are designated as important to safety described in Appendix A, which support the HB ISFSI, Radioactive Materials Transport Packages, and the HBPP Unit 3, and as defined within other portions of this section. Sufficient records shalJ be maintained to furnish evidence of both the quality of items and activities affecting quality and to meet applicable code, standard, and regulatory requirements. The records include ali documents referred to or described *in the HBQAP or required by implementing procedures such as operating logs, maintenance and 1)1odification procedureS, J:elated inspection results, ~nd reportable occurrence~; and other records*

required by the applicable License, Technical Specifications and Code. of federal Regulations. In addition to the records of the results ofreviews, designs, fabrication 7 instalJation, inspections, calibrations, tests, maintenance, surveillances, audits, personnel qualification, special process qualificatio~, and material analyses for PG&E quality-related *activities and SSCs that are importanno safety, those of vendors, supplierS, subcontractors, and contractors shall also be maintained.

A management control system for the collection, storage, and maintenance of completed quality assurance (QA) records shall be maintained. This records management program shall be designed and implemented to assure that the QA records are complete, readily retrievable when needed, and protected from damage or destruction during storage by fire, flooding, theft, environmental conditions, or other causes.

QA records stored electronically will follow the guidance for electronic records management given

• in the Nuclear Information and Records Management.Association (NIRMA) technical guidelines, TG ll-1998~ "Authentication of.Records;" TG 15-1998, "Management of Electronic Records;

TG 16-1998, "Software Configuration Management and Quality Assurance;" and TG 21.:.}998,

~'Electronic Records Protection and Restoration. QA records Will be stored on electronic media (thatis; optical disk, magnetic tape, network array, etc.) meeting the requirements of the NIRMA guidelines. Alternately, records stored on optical disks may meet the requirements of Generic Letter88-18, "Plant Record Storage on Optical Disk," dat~d Octob~r 20, 1988, Information

• Systems will detennine the appropriate electron'ic media. Regardless of the electronic media selected, the pro~ess must be capable of-producing legible, accurat¢, and complete records dqring.

the requited retention peri'od.

Backup copies of in-process *electronic r:nedia records will be maintained in nnlltipJe, physicaJiy-independent electronic loca.tions. Backup ~opies of QA records in electrqnic media will be maifitained in multiple, physically-independent electroni.c locations until such time as images of

• thesl'(* records are cre&t(!d; copied, .and verified on tWo cop*ies of an ~ppropriate electronic storage media. The two copies will then he *stored In separate physica:i iocations. File legibility verification will be completed on.all QA records stoted on electronic media byeither visually*

ve.rifying the file legib,il ity or by eJectronicaJ ly verifying e~act bi11ary file transfer. Periodic media inspections to .monitor image degradation. will be conducted in accordance with the NIRMA guiqelines or m*edia manufactUrers' recomrn*endations. These pedodic inspections shall be documented. QA records stored on electronic media wili be refreshed or copied on to new media arid subsequently verified* if.the projected Hfetime ofthatrnedia does not exceed the

.ret~ntion period ofthe recor~s $tqred on .that medi~. Thes.e req~itel'n~i1ts m¢et th~dnte11t of Getieric tetter 88-18.

D~~ile.d recqrds -for items or activi.ties shall be ~pecified by instn1ctjpns, pmcedures, drawi.f\gs,. or Specification or other documents .that prescribe the item or activity and shait be generated hy the otganizatibn responsible for the item or a:c.tiv.ity inc !tiding PG8iE and Q.on.~PQ&E organizations:.

L-4 Attachment 2.1 Rev. 31 Page 29 of 52 Each department generating QA records is responsible for transmitting those t*ecords to the records processing organization for archival purposes.

All records shall be assigned a retention period in conformance with Title 10, Code ofFederal Regulations, other applicab]e codes, standards, and specifications.

HB ISFSI Records shall be classified as lifetime ot nonpennane.nt. The following records shall be maintained as required for the HB ISFSI: *

1) Radiation proJection program and surv~y recor<is

2) Records associated with reporting defects and noncompliance

3) Records important to the decomn:ti$Sionihg of the .ISFSJ ( 10 CFR 72.30d)

4) Records of changes to the physical security plan made without prior NRC approval

5) Records of changes, tests and experiments, and* ofch~ges to procedures described in the HB ISFSI FSAR Update pursuant to 10 CFR 72.48 . .

6) Records showing receipt, inventory, location, disposal, acquisition, and transfer of spent fuel

7) A copy of the current inventory of spent fuel in storage at the HB ISFSI

8) A copy ofthe current material control and accounting procedures

9) Other records required by-license conditions or by NRC rules, regulations or orders

10) Records of tne occurrence and severity of impo*rtant natural phenomena that affect HB ISFSJ desi~

• 1 J) QA records (inc]uding. records pertaining to the design, fabrication, erection, testing, maintenance, and use of structures, systems, ~nd components importanlto safety; and results .

of reviews, inspections, tests, audits, monitoring ofwork performance, and material analyses) 12} A copy of the current p])ysical security plan, pius any s'ttperseded portions of the plan

13) A copy ofthe current safeguards contingency plan procedures, plus any superseded portions of

• the pr0cedur~s * . *

14) Operating records; including maintenance, alterations or additions made

15) Records of off-norma) o~currences and events

.16) Environmental survey records

17) Records of lndependent Safety Reviews

18) Records QfNuclear Safety Oversight Committee (NSOC) 19).Records ofempioyee qualifications and certifications

20) Record copies ,af:

• HBlSFSl.FSARUpdates

• Reports ofaccidental.criticaHcy or lossofspecial nuclear material"

• Material st~tus reports

• Nuclear ,rrta~e.rial. transfer reports

• • Reports of pre~operational test acceptance criteria and resolts

• Procedures

• Environmental Report

• Emergency P Jan 2.1) Cops.truction Recoi-ds

22) Records *ofe~ents :associated with radioactive *releases HJ;3PPUpit 3 Records sh()IJ be clas.sified as lifetime or t}onperman~nt. Th.e following reqords shall be. ma*intained asrequl"red for the HBPP Unit 3:

• All record-s and logs relative .to the fo II owing areas: shall be retained for at least 5 years:

1) Records and logs of normal SAFSTOR operations

L-4 Attachment 2.1 Rev. 31 Page of52

2) Records and logs of principal maintenance activities, including inspection, repair, substitution, or replacement of principal items of equipment ~escribed in the Technical Specifications *

3) Reportable Occurrence Reports .

4) Records of periodic checks, inspections, and calibrations. perfonned to verifythatsurveillance.

reqlikements are being met

5) Records of radioactive shipments

6) Records ofs.ealed source leak tests and results

7) Records of the annual physical inventory of all source material of record

8) Records of changes made in operatiJ!g procedures All records relative to the following areas shall be retained for the duration of:SAFSTOR and decommissioning: *

1) Records and prints of changes made to the Plant

2) Records ofplant:radiation and contamination*surveys

3) Records of offsite environmental monitoring surveys

4) Records of radiation *exposure for all plant personnel, including all contractors and vishors to the plant, in accorda~ce with 10 CFR 20

5) Rec*ords of radioactivity in liquid and gaseous wastes released to .the eQvironment

6) Records of training and qualification for current members of the plant staff

7) Records of Independent Safety Reviews (ISRs)

8) Records from Nuclear Safety Oversight. Committee (NSOC) meetings and reviews

9) Records ofQuaJity Assurance activities ~quired by the HBQAP I 0) Records ofreviews performed for changes .made to procedures* or *equipment~ .or reviews of tests an:d exp.erim.ents porsu.antto theDefueledSafety Analysis Report(DSA.R)

• 11) Recotds of reviews performed for changes made to the Offsite pose Calculation Manual and

• thePto,cess Control Program

• L-4 Attachme1it 2.1 Rev. 31 Page 31 of 52 18.0* AUDITS The requirements for Audits apply to those SSCs and associated activities that are designated as important to safety described in Appendix A, which. support the.HB ISFSI and the HBPP Unit 3 or as defined within this section. The adequacy and effectiveness of the Quality Assurance (QA)

Program shall be continually monitored through a comprehensive system of internal and supplier audits. The audit system implemented by the QV organizati'on includes all aspects of the HBQAP.

The audit system shall:

(1) Verify, through exaQ1ination and evaluation of objective evidence, that ~his HBQAP has been implemented as required (2) Identify any deficiencies or nonconfonnances in this HBQAP (3) Verify the correction of any identified deficiencies or nonconformances (4) Assess the adequacy and effectiveness of this HBQAP (5) Assess frequency and effectiveness of use of supervisors as design verifiers to guard against abuse

• A comprehensive plan fOJ::the auditsystem shall be established and documented. Audit frequencies are determined by a performance-based evaluation plan. This p1an uses assessment indicators to identify arid ..schedule audits based on performance results and importanc.e of .the activity relative to safety. The plan shall identify the scope of individual audits that are to be perfonned the aspects ofthi~ HBQAP covered by each audit, and the schedule for pe1fonning audits. The audit. system plan shall be reviewed at least semiannually and revised as necessary to assure that coverage and the schedule reflects current activities HBPP Unit 3 and HB ISFSI activities are being accompli~hed in accordance with applicable requirem~nts. Other associated activities inCluded as part of the audit program are: indoctrination and training programs; the qualification and verification of implementation ofQA programs of corttr~ctors and suppliers; interface control among the applicant and the principal contractors; audits by contractors ~nd suppliers; corrective ~ction, calibration, and nonconformance control systems and HB. ISFSI commitments.

Auditors shall be independent of direct responsibility for the performance ofthe activities that they audit, have experience or training commensurate with the scope and complexity of their *audit responsibility and be qualified in accord~mce with USNRC'Regulatory Guic{e 1.-146 (Augus.t, 1980), "Qualification .of Quality Assurance Prqgram Audit Personnel for Nuclear Power Plants."

Auditing shaH be initiated as eady in the life of an activity as is practicable and consistent wit\1 the schedule for accompiishing'the activity. 1n any case, auditing shall be initiated early enough to as.sute that this HBQAP is effectively implemented thro1,1ghout each activity~ lndividua'l audits shall be regularly scheduled Oil the basis of the statQ~ and importance of the activities, which they address.

For audits, other than*those with scheduled frequencies that are* is mandated by regul~tion have a grace .Period of up. to 90 days, which may be utilized when the orgei1cy of other priorities makes meeting the specified sc1w.dule dates impracticaL For aud~t activities deferred by using a grace period; the next scheduled dUe-date shali b.e based on the original schedule due date, but may not excee<t the origin~ I d~e date plu~ 90 days.

L-4 Attachment 2.1 Rev. 31 Page32 of52 Audit rep011s shall be prepared~ signed by the Audit Team Leader, and issued to respons~ble management of both the audited and auditing organizations.

Audits ofHB lSFSl activities shall be performed under the cognizance of the NSOC as defined in Appendix B. The NSOC shall re*port to and advise the CNO on the tesults of the audit program{

Audits are reguhirly scheduled on~ formal audit schedule prepared by the QV organization. The audit schedule is reviewed regularly by the QV Director and the schedule is revised as necessary to assure adequate coverage as commensurate with activities and past perfonnance. Audits are pelformed in accordance wjth approved audit plans. Such audits may be augmented by other QV assessments aod independent inspections, using approved procedures. Additional audits may be performed as requested by the CNO, the HB Director, or the QV Director.

The following HB ISFSI areas shall be audited.at least once per 24 months or more frequently (iS performance dictates:

1) The conformance of J-IB ISFSI operations to provisions contained within the applicable Technical Specifications and L~cense.

2) The performance, training, and qualifications of the entire HB ISFSI staff.

3) The res:ults of actions taken to correct deficiencies occurring .in HB lSFSI equipment, structures, systems, or method of operation that affect nuclear safety.

4) The performance of activities required by the HBQAP to meet tbe criteria of 10 CFR 50 Appendix. Band 10 CFR 72 Subpart G.

5) A representative sample of routine HB ISFSl procedures that are us~d more frequently than every tw() years. This audit is to ensure the acceptability of the procedures and to verify that the* procedures review and revision program is being implemented effectively.

. 6) *The performance Qfactivities required to be audited by ANS~3.2/ANS1 Nl8.7-1976, Section 4.5.

7) Review of design documen.ts and process to ensure compliance withSecti.on 3.0.of:this HBQAP (i.e., use of supervisors as design verifiers). In addition> QV shall sample and review specifications and design drawjngsto assure that the documents, ate prepared~

reviewed, and approved in *accordance with PG&E procedures ;~nd that the documents contain the. necessary QA requirements, a.cceptance requirements, and quaiitydocumentation requirements.

8) QV shall audit the departments that qualifY personnel and pro~edures. to assure that the*

process qt1a_lification activity, records, and personnel meet the applicable requirementS. They

• shall also audit the organizations impiementing special. processes .to provide assurance* that the processes are carried out in accordan9e with approved procedures. by qualified *personnel using qualified equipmenl and ~haf required records ~re properly maintained.

9) The~adiatioP. Protection Prog~~m i.n ~ccqtdance with lQ CFR ~0.

10) T:be Emergency Plat1 an<l Jn1plemendng pi*ocedures.

L-4 Attachment 2.1 Rev. 31 Page 33 of 52 II) The Security Program in accordance with 10 CFR 73.51(~)(12).

12) The HB ISFSI Access Authorization Program in accordance with 10CFR73.56 (n)(l)

The following HBPP Unit 3 areas shall be audited at least once per 24 months or more frequently as performance dictates:

1) The conformance to provi*sions contained within the TechnicaJ Specifications and applicable license conditions. *

2) The performance, training, and qualifications of the entire Unit 3 staff. *

3) The results of actions taken to correct significant deficiencies occurring in Unit 3 equipment, structures, systems or methods of operation.

4) The perfom1ance of the following activities required by the Quality Assurance Program.

a) Quality Assurance Program b) The Radiation Prot~ction Program c) Radiological Effluents Program d) Radiological Environmental Monitoring Program e) Radioactive Material Packaging and Transportation f) Radioactive Waste Processing and Process Control Program

5) The Emergency Plan and implementing procedures.

6) The Fire Protection and LQss Prevention Program and Implementing Procedures. The audit team wi II. include qualified offsite licensee personnel or an outside consultant.

Supplemental audits (or independent assessments) shall be performed as authorized by the QV Director based on the following considerations:

• \Vhen significant ~hanges are made in functi<mal areas of the HBQAP such as significant reorganizations or procedure changes.

• When it is suspected that the quality of an item or activity is in jeopardy ciue to deficiencies in theJIBQAP.

• When a systematic, independent assessment of program effectiveness is con~idered necessary.

• When supplemental audits or.assessments are necessary to verify implementation ofrequired corrective action.

L-4 Attachment 2.1 Rev. 31 Page34 of52 Management of the audited organization shaiJ review the audit report and respond to any quality prob]em reports, investigate any significant findings to identify th~ir cause and determine the extent of COJTective action required, including action to prevent recurrence. They shan schedule

• .such corrective action and also take appropriate action to assure it is accomplished as scheduled.

They shall respond to QV regarding each significant finding stating the root cause, immediate action taken,.and the.corrective action taken or planned to prevent recurrence. Such responses may be documented directly within electronic databases used for the corrective action p~ogram.

QV shall review t.he written responses to" all audit findings, evaluate the adequacy of each response, assure that corrective action to prevent rec\,lrrence is identified and tak~n for each significant finding, and confirm that corrective action is accomplished as scheduled.

Audit records shall be generated and retained by QV for all audits.

L-4 Attachment 2.1 Rev. 31 Page 35 of 52 .

APPENDIX A IMPORTANT-TO-SAFETY STRUCTURES, SYSTEMS AND COMPONENTS The pertinent quality assurance requirements of IOCFRSO Appendix B, 10CFR 71 Subpart Hand 10 CFR 72 Subpart G wiJI be applied, as a minimum, to all quaUty aGtiviti~s affecting the important to safety structures, systems and compon*ents (SSCs) associated with spent fuel storage and transportation package that are listed below.

NOTE*

There are no . HBPP safety related SSCs remaining at the site.

The quality classification ofNRC LiGensed HB lSFSl Dry Fuel Storage Components and Transportation Packages must be made by the NRC Certificate Holder. The Certificate Holder is responsible for design and licensing controls for these components under their NRC approved Qua.lity Assurance Pr()gram. PG&E lltilizes these types of co_rnponents .and packages under the provisions of a NRC 'General ~icense for Radioactive Material Transportation Packages (10 CFR

• 71) and Soent Fuel and GTCC S toratZe ( 10 CFR 72).

IMPORTA.NT TO SAFTY AS DEFINED BY 10 CFR 71 AND JO CFR 72 A. Dry Spent FueJ and GTCC Storage (10 CFR 72) sse Quality *Design/License Category(a) Responsible Mu1t.i~Purpose Canister A :PG&E Fuel 'Basket and Basket Spacers A P,G&E Damaged Fuel Container A PG&E Hi-Star 100 HB Overpack A PG&E Transp.orter Uft Links A PG&E GTCC W ast.e .container A .PG8iE Hi-Star HB GTCC Overpack A *po~E fiB lSFSI Storage VauJt<cl B PG&E Fuel Spacers B PG&E Transporter Connector Pins B :J>G&E I:I.~Hum 'Fill GMt~>> B PG&E Lid Retention DeviGe B P..G&E C~sk Transportertt>> B llG&;E Proce~s W.aste <;ontainer B *pG&E B; Transport of.SpentFueland GTCC'Waste (lO*CFR 71) sse Quaiity Design/License Responsible C~te2ory(a)

MtJ_l{i-Pt,J.fPOS~ c~mis~er A HpJ.tec International Fuel. Basket and Basket Spacers A 'I{oltec International Damaged .Fuel Container Holtec International Hi-Stat 100 HB Overpack A lloltec Interhationar A Jtoltec lntemationat Hi*-Star liB_ OTCC *overpack A . Holtec lnte~national Fuel Spacers* B

L-4 Attachment 2.1 Rev. 31 Page 36 of 52 C. Radioactive Material Transport Packages- Greater than Type A Shipments (10 CFR 71) sse Q~ality. Design/License Responsible Category(A)

TBD* TBD* NRC Certificate *Bolder*

• Radioactive Material Transport Packages subject to the provisions of 10 CFR 71 t Subpart C, General Licenses are "Important-to-Safety" and subject to the applicable requirements of the HBQAP. sse and Quality Category are as det~tinined by the NRC Certi"flcate Holder for-the cask chosen for the shipment.

NOTES:

1. See Holtec International ill-STORM Safety Analysis Report (SAR) and associated spepifications for additional classifiGation infonnation.

2.

• See Ho1tec lntemational ffi.STAR Safety Analysis Report and assbciated specifications for additional classiflcation.infonnation.

3. For the defmition of Quality Categories A, Band C reference NUREG/CR-6407 (a) Major cask system components are Jisted acc.ording to the highest.QA category of any subcomponent comprising the.major component. The safety .classification of the subcomponents and the determination of the ITS category of each, item is adn1inistrativ.ely ~ontrolled by PG&E via design and procurement control procedures with input from the Storage disk vendor.

(b) Purchased commerCial grade and qualified by testing prior to use.

(c) HB I$FSI storage vault Jid and lid closure bolts are classified as ITS Category B except for* beyond design basis :soil structure interaction seismic events. For these postulated SSlseismic events~ the HB ISfSI storage vault lid and lid closure bolts -are classified as NITS stnce they ~re oot relied. upon in the seismic accident analysis ..

to Refer Section 3.2.4 and PG&E Lettet HIL-05~007; dated June 3'? 2005, for details ofthe classification *

.(d) The storage vault drainage _pipe is cJassified as NlTS because the drainage system is pn]y one -of S,eYeral *design. fei:ltur_es relied .upon to e*nsure .a~f"eql,!a.te performance of the in cask and storage vault system in the event ofstanding water the vault cells.

Details of the assessment performed are provided in PG&E RespcinsetoNRC

. Qu:estion 5-10 jn PG&E Letter HIL--04-007; <.lated Qc;tober l, 2004,

lr4 Attachment 2.1 Rev.31 Page37 of 52 APPENDIXB ADMINISTRATIVE CONTROLS 1.0 PROGRAMS AND PROCEDURES

1. In addition to those procedures required by HBQAP Section 5.0) the. procedures and programs listed below shall be establish'ed and controlled to support the HBPP Unit 3 and HB ISFSL These procedures and changes thereto shall be reviewed by Independent Safety Review and approved.by the HB Director (or designee) prior to implemen~tiont except as provided in items 2 and 3 below:

a) Normal startup, operation, and shutdown of systems arid components required during SAFSTOR b) Actions to be taken to correct specific and foreseen potential malfQnCtions of systems or components c) Actions to be taken during emergency conditions involving unplanned releases of rad-ioactivity d) Abnormal and emergency operation of all systems and components required to maintain the SAFSTOR condition of the Plant e) Surveillance activities required to demonstrate compliance with the ~pp Unit 3 Technical Specifications t) Calibration *of-instrumentation used to demonstrate compliance with HBPP Unit 3 Technical Specifications

• g) Shipping and disposal of radioa~tive material~

h) Process Control Program i) Fire Protection Program implementation j) Activities referenced in the HB ISFSI Technical Specifications.

k) Activities related to monitoring ofrad.ioMtive effiuents and the enviro.nment.

I)

• Secllrity, maintenance, or administrative procedur~s related to any ofthe above m) Other procedures deter~ined by the QV Director or .I-Il3 Director to be significant to the liBPP Unit 3 or HB ISFSI. .

2. Tetnpcmtry Procedure Changes Adnilriistrative ,controls shall be established that provide the methods by which temporary changes 9an be made to approved ptocequres~ incJuding the design_a~ion of those persons authorized to approve such changes. Temporary changes that dearly do not change the intent ofth~ ~pproved procedure from the standpoint of nuc1ear safety may be approved *by two members ofthe plant management staff~ Such changes shall be doct~rnented and, if apprqpdate, .in.corporated in.to the next revision of the affected :procedure. Administrative

-~ontrots*forapptov~i and timely notification or training of personnel aff~ct~. bY the temporary change. shaH be .implemented. The change is documented, revtewed'as: described *above, arid approved'by the ;Jppropriate a:pprov.al authority within 14 days of:implementation.

L-4 Attachment 2.1 Rev. 31 P~ge38 of52

3. Emergencies Not Covered by a Procedure

• ln. the event: of an emergency not covered by an approved. procedure designated site personnel shall be instructed to take action to minimize personnel if!jury and damage to the facility.

4. Periodic Review The procedures required by HBQAP Section 5.0 and HBQAP Appendix B Section 1.0, the Emergency Plan, and the Security Plan, shall be subject to biennial review. This shaH be controlled hy ~dministrative. procedure~ ~nd shall includelndepen~ent Safety Review. Resu!ts and recommended. changes shall be submitted to the f(B Pirector. -

2.0 INDEPENDENT SAFETY REVIEW An Independent Safety Review shall be a thorough revieW condu_cted .by one or more qualified Independent Safety .Reviewers. Persons perfonning these reviews shall be knowledgel;lble in the subject area being revi'ewed. Independent Safety Reviews must be completed prior to implementation ofproposed activities.

1. Independent Safety Reviewers shall be individuals without. direct responsibility for the performance of these activities under review. These reviews may be from the same functionally cognizant organization as the individual or group performing the original work.

2. Independent Safety ,Reviewers shall have at least 5 years. of professional experience and either a Bachelor's Degree in Engineering or the Physical Sciences or shall have equivalent ql.lalificationsin accordance with ANSI 18.1-1971. the fiB. Director-shall docum¢nt the appointment 9f lndependent Safety Reviewers.

3. The following subj~c;ts shall be independently reviewed by a qualifi.ed Independent Safety Reviewer: * *

• a) £v~luations*(i0 CFR 50.59 or 10 CF'R 72.48* assessments which do not screen out) for changes in the fa_cility .or HB ISFSI as desc.dbed in the Defueled Safety Analysis R~pprt (DSAR,) for the- J:IB.:PP Unit 3 or the HB lSfSI Final Safety Analysis R,eport(FSAR) being, appliedtotheHirlSFSI;

• b). Evaluations (10CFR 50.59 or .10 CF:R-72.48 assessments which do n9tscreen out)(or proposed changes to the *Emergency Plan or Security Plan; c) =Changes in procedures as .described in the DSAR or FSAR, or .test~ or experiments not described:in*theDSARor.FSAR to verify'that such actions-do not Involve a change to-the Technical Specificf:l.tions.orwill not require prior NRC approv~l as defin~d in tO CF:R 50.59 or 1tYCFR.72A8; . .

d) Proposed changes to the _programs and procedure~ r~quired by the BBQAP in Section 5.0, or AppendiXB. Sectio.n 1:.0~- to verify that such changes. do not invo.lve.a change to the .

Techniqal Specificatiol):S*orwil.l not requite prior NRC approval ~.s defined in lOC.FRso;s9 or-1 OCFR72.48;

.e) Proposed te$ts and. ~xperim~n~s ~etermined by t,he HB Pir~itor to h.~v~ ri\lclear safeiy *

• significance; *

• L-4 Attachment 2.1 Rev. 31 Page 39 of 52 f) Proposed changes or modificat~ons to the ITS SSCs identified in Appendix A; and g) .Proposed cbanges to the HBPP Unit 3 or HB ISFSI Technical Specifications or Technical Specification Bases.

h) Reviewing, pti()r to approval, new proc~dur~s.used to handle heavy load~ in exclusion areas and changes directly related to methods and routes used to handle heavy loads in exclusion areas.

4. Independent Safety Review sh~ll also assess and ensure .that appropriate cross-dis~ipline review has been app1ied _to th~ document under review.* *

5. Independent Safety Review shall recommend in writing to the appropriate approval authority, approval or disapproval of the items considered above. *

6. Independent Safety Review shall render .determinations in whether each of the items consider~d ab()ve require prior NRC approval.

• 3.0 NUCLEAR SAFETY OVERSlGHT COMMITIEE (NSOC)

The QA Program also includes an independent revi*ew function, implem¢nted by Nuclear Safety-Oversight Committee (NSOC). This function provides an independent review of

• HB lSFSI o.r Unit 3 changes, tests, experiments and procedures, which constitute a change to the HB ISFSI as described in the HB ISFSI FSAR, or the HBPP Unit 3 DSAR. In additio.n, the-NSOC will verify that reportable ev_ent$ are irive~tig~ted ln a tiro.ely manner*

and :corrected In a manner that reduces the probability of recu1Tence of such events; and detect trends that may not appear to a :day-t~day observer..

The NSOC Chair and members s.hall be appointed in writing by the CNO. The individuals a_ssign¢~ t~sporisibilityfor NSOC reviews sh~ll be qualified in. specific disciplines. These ind-ividuals* shall collectively have the experience and competence required to review activities in the following are.as:

(1.) HB ISFSI and Unit 3 SAFSTOR operations (2) Nucle~r engineering 0} *cherrtistry and :radiochetnistcy (4) Metallurgy (5) Nondestructive testing (6) m$tnent arid contro.l (7) Radiolog_ical safety (8). -M~haoj_ca] and electrical engin~ering

{9) Administrative controis (1<0 Quality assurance practices

,(ll) Oth_er*appr:o*priate fiel.d$

L-4 Attachment 2.1 Rev. 31 Page 40 of 52 NSOC shalJ report to and advise the CNO on those areas of responsibility specified in the sections below.

Composition - NSOC membership shall be comprised of site represent~tives and offsite members. Membership will normally include the QV Director and four members. The NSOC Chair shall have a minimum.of6year:s of professional level managerial experience in the power field and NSOC members shall have a. minimum of 5 years of professional level experience in the power field.

'(he NSOC Chair and all ~embers shall have qualifications that meet or exce~d the requirements and recommendations of Section 4.7' of ANSJ/ANS 3.1*1978.

• An individual may possess competence in more than one specialty area.

Consultants: Consultants shall be used as determined-by the NSOC Chairto provide expert advice t_o NSOC.

Meeting Frequency: NSOC shall meet at lea*stt,once per quarter.

Qu.ont~: .A quorum ofNSOC is.necessary for the performance of the NSOC function required by the QA Program. The quorum shall be a majodty of the members, and shall include the-Chair (or appointed Vice-Chair).

Review: NSOC shall review:

(1) The evaluations for:* (a) changes to procedures, equipment, or systems, and.(b) tests or experiments oomplete(J under the provision o.f I 0 GfR 50.59 or 10 CFR 72.48, to verify that such actions did not require prior NRC approval (2) P.rpposed change*s to procedure*s, ¢quipment, or systems, -that require prior NRC approval as defined in 1.0 CFR50.59 or. 10 CFR 72.48 (3) Proposed tests_or experiments that require prior NRC approval as defined in IO CFR 50.59 oJ; 10 CFR 72A8 .. .

( 4.) Propnsed changes to the HB ISFSI or Unit 3 Technic.al. Specifications or Hcense (5) Violations of codes, regulati9ns; order~, T~chnical Specific~tions,..Iicense requirements, or of internal procedures *or instructions having nuclear safety sign ifi¢ance

.(6) Significant operatingabnorm~lities or deviations frorn normal ~d expected

• perfo.tmance of B1l iSFSl or Unit. s* equipment that affect nuclear s.afety (7) AJI reportable ev~nt_s_

{8) All recognized indic~tions of~ unanticipated deficiertyyjn _some.a~pect ofl1Jl

_ ISFSl or Unit 3 design or operation ofimportant-to~safety SSCs that' could *affect tiucle*ar ~afety,

.(9) Significant accidental, unp~anned; or nncontroUed rqqio~ctive relea~es.

{10) The performance :of.the corre*ctive actlon system.

(11) Maintenance and surve.illance testing experience to .enstire: safe and efficient_

m~intenance ofthe XSfSI ~nd:UnHJ, Jlnd to determine i:f~h~nges fo ~quipment or_procedures are needed

L-4 Attachment 2.1 Rev. 31 Page4l of52 (12) Internal and external experience information related to the safe storage of irradiated fuel that may indicate areas for improving facility safety (13). Any other matter involving *safe operation of the HB ISFSI or Unit 3 that the .

quality verification director deems appropriate for consideration, or which is referred to the director by organizational units NSOC rnay delegate reviews of selected topics such as changes processed t.mger *1 0 CFR 50.59 and J 0 CFR 72.48to QV. The appropriate NSOC subcommittee will consider QV's reviews of those topics in their meetings.

I o .. -

Records: *Records.ofNSOC reviews and activities shall be prepared; approved, ahd distributed as indicated below:

( 1) A summary report shall be prepared, approved, and forwarded to the CNO and HB Director..

(2) Summary report shall include evaluation, recommendatio11s, and disposition of the corrective action to prevent recurrence.

(3) Mirm.tes of each NSOC meeting_ shall be prepared, approved, and forwarded to the CNO within. 30 *days followi~geach meeting The NSOC's responsibility shall includ~ the Audit Program. NSOCaudit~ shall be perfonned of the activities and at the associated frequendes defined in Section 18 of this HBQAP.

Written records ofreviews and audits shaH be maintained under the provisions ofSection 17 .of the HBQAP. As* a minimum, these records shall include:

{1) Results of the activities conducted under the provisions of Section 3.0 of the HBQAP Appendix B; (i) Recommendations to the. management of the ,audited *organi'zation; (3) An ~ssessment ofthe safety significance of review or audit findings; (4) Documentation of rcview.s comiucted tu~der Section 3.0 of the HBQAP Appendix B,

, and *

• (5) Determination of whether each item considered under Section J.O of the HBQAP Appendix B reqQir~ ptior NRC appr:oval.as defined in 10 CFR 50:59 and J0 CFR 72.48. .

4*.0 RAl)IOLOGI~A.L MONITORJNG Admlnastrativecontrols for rad.iolog'ical ~onftoring. of gaseous and liquid :eftluents and the envjronm~ht _sh~ll be conttolled and implemertted to ensu:re c.onfonnanc.e with USNRC Regub1tory Guide 4. 15,(December 1977);.

5.0 TECHNICAL SPECJFICATION ACTJVITIES ln addition :to th¢* applicable qu_ality as$1Jtanc.e requ'itetn.ent$ sp,eclfied in the I-mQAP, Technical Spe~iijc~tion.~ctivittes ~h~.U be controlled in,~ccord~n~e with the .Lirnitillg*Cpndition~ for Operations (LCO) and Surveillance Requirements (SR).

L-4 Attachment 2.1 Rev. 31 Page42 of52 6.0 OFFSITE DOSE CALCULATION MANUAL The ODCM shaH contain the methodology and parameters u_sed in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation -of gaseous effluent monitoring alarm aod trip setpoints, and in the conduct of th~ radiological envitonmental monitoring program; and The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the inform~tion that should he included in the Annual Radiological Environmental Monitoring Report, and An.nual Radioactive Effluent Release*

Report, required by Appendix B Section 8.2 and 8.3 of this HBQAP, respectively.

Licensee initiated changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This doc*umentation shall contain:

i) sufficient infonnation to support the Ghange(s) together with the appropriate analyses or evaluations justifying the change(s), and ii) a d~tetmination that the change(s) wiJJ maintain the level of radioactive effluent control required by *1 OCFR 20.1302, 40CF-R Part 190, 10CFR 50.36a and Appendix I -

to 10CFR 50, and not adversely impact the accura¢y or reHability of effluent, dose, or setpoint calculations; 2.. ShaH become effective after review and approval of the Plant Manager; and J. Shall be s.ubmitte.d to the Commission in the form of a complete, legible ~opy of the entire ODCM as a part of or concurrent with the Radioactive Effll]entJtelease Report for the period of the report in wh lch any change in the ODCM was made. Each change shall be identified by markings in the margin o.fthe affected p*ages_, cle'lfly indicafing_the area ofthe page: that was changed, and shall indicate the date (i.e., month and year) the change was implel)ier)t¢d.

7.0 RADIOACTIVE EFFLUENT CONTROLS PROGRAM This program conforms with fOCFR s*0.36a for the control of radioactive effluents and for maintaining the doses to m¢mpers ofth~ public from radioactive *effluents a~ low as reasonably achievable (ALARA). The program()) shall be contained inthe:ODCM, (2) shall be.

implemented by operating' proc~ures~ an*d ,p) shall include rem~dial actions to be taken whenever the program limits are exce.eded. The: program shaH Includ-e the followjng eiements:

1. *Limitadons on the operability of radioactive gaseous monitoring: instrumentation 1ncluding

. s*urvelllanc~ tests and -setp'oint d~t¢rmiJ"Httion *in *acco_rdance with the methodology in the

• ODCM, . .

2. Limitations on the instantaneo.us. (average over a one-hour period) concentrations of radioact.iv~- material tel.eas~d Tn l iqWid. effluent$ to Humboldt :B:ay ¢Qnformiog 'to t~n fhtl'es the eftluent=concentratlon limits ofl OCFR Part'20, 1\ppend 1x"B; Table*2, Column 2,

3. Monitoring, sampling, and analysis of.r~dfoactive liquid and gaseous effluents in accordance w.im lQCFl\ 20.J 302 and wjth ~he. m.ethopology and* pnramet~r$ in the QDCM, *

4. Limitations on the annualan*9 quarterly* dos~s or*dos~ c*omrni"tm~nt t_o:a m.:emh~r ofthe public. -from. radioac.tive. materials in llquid effluents releas_ed to Humb.oldt Bay conforming to th.¢ .dos.e design *objecliV¢$ ofAppen_di~ l to 1OC:FR Part: 50'- -

L-4 Attachment 2.1 Rev. 31 Page 43 of 52

5. Detennination of cumulative and .projected dose contributions from radioactive effluents fm the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at leasfevery 31. days,

6. Limitations on the operability and use of the liquid effluent *tr~atment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment confonning to the dose design objectives of Appendix I to IOCFR Part 50,

7. Limitations on. the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary shall be established as follows:

a. For tritium and radionuclides in particulate form with half-lives greater than 8 days:

less than or equal to a dose rate (averaged over a one-week period) of 1500 mrem/yr to any organ.

8. Limitation on the annual and quarterly doses to a member of the public from tritium and all radioriuclides in particulate form with half-lives greater than 8 days in gaseous effluents re.leased to areas beyond the site boundary conforming to the dose design objectives of Appendix I to 10CFR Part 50) and

10. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40CFR Part 190. .

L-4 Attachment 2.1 Rev. 31 Page 44 of 52 8.0 REPORTING REQUIREMENTS The foHowing reports shall be submitted in accordance with 10CFR 50.4.

8.1 Occupational Radiation Exposure Report An annual report shall be made of p~rsonnel exposure, in accordance with the requirements of 10CFR Part 20.2206. *The r:eport shall be s.ubmitted by April30 of each year.

8.2 Annual Radiological Envirmtme;nta*l Monitoring Report The Annual Radiological Environmental M9nitori.ng Report covering the oper~tion of the unit*

during the previous calendar year shall be submitted by May I of each year. The report shall inClude summaries, interpretations, and analyses oftrends.ofthe results ofthe Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the ODCM, and in l OCFR 50, Appendix 1, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiological Environmetital Monitoring Report shall include the results of analyses of radiological environmental samples:and of environmental radiation m~asurements taken during the period p*urs\lantto the quality related .locations specified in the table and figures in the ODCM, as well as summarized and t~Qulated results of these analyses and measurements. Iri the event that some individual results.are not available for inclusion with the report, the report shall be submitted noting and ~xplaining the reasons for the missing results. The missing data shall be submitted in the next annual report.

8.3 Annual Radioactive Effluent Release Report The Annual Radioactive Effluent Rel~ase*Report covering. the activities:ofthe unit in the previous year shall be submitted prior to April 1 o(each y~ar in ac_cordance with 10CFR SOJ6a. The report *shall include a summary of the *quan.titie~ of i'Fldioactiv.:e liquid and gaseQtiS efflu~nts and solid waste released from the unit. The material provided shall be consistent with the objectives outline.d in the QDCM and in confon.nance with. 10CFR 50.36a and *l OCFR Part 50, Appendix 1,Section IV .B.l. .

L-4 Attachment 2.1 Rev. 31 Page 45 of 52 AP*PENDIXC CURRENT REGULATORY .REQUIREMENTS AND PG&E .COM'MITMENTS PERTAINING TO THE QUALITY ASSURANCE PROGRAM The* Quality Assurance* Pn:lgram forthe Humbol.dt Bay HB ISFSI, described in the HB QA Plan, program directives, and administrative procedures complies with the requirements set forth .in the Code of Federal Regulations. In addition, it complies with the regulatory documents and industry standards listed below.

Changes to .this: list are not ma*~e without the review and concurrence of the QV Director.

R~g. De1te 'Stand~rd Rev. Tifi~/Su bject Exeeptions Guide$ N<>.

(?. G.) 6/72*: ANSIN45.2 1971 Quality Assurance 28 . Program

.ReqUirements for NUclear Power Plant$

1.3~ 5/77 ANSI N452.2 1972, :Quali{9-A$_surance Warehous~ personnel will normally _visually scrutinize-Incoming shipments for

_Requirements for damage of the types li$ted in Section 5.2.1, this activity is not necessarily P*ackaging, performed prior to unloading. Separate documentation of the shipping damage is Shipping, .Receiving, not n*e~ssary. Re.lease of th.e transport ag*ent after unloading and the signing for Storage, and receipt of the shipment provides adequate documentation of-completion of the Hanoling of ltem*s .for shipping oamage inspection. Any dam*age noted will be documented and Water-Cooled dispositioned.

Nuclear Power Pla.nts

• Persons performing this visual scrutiny are not considered to be performing an inspection function :as defined under Reg. Guide 1. 74; therefore they do not require certification as an inspector under Reg. Guide 1.58.

1.39 *9/77 *ANSI N45.2.3 1973 Housekeeping Housekeeping zones established at the power plants differ from those described in Requirements. for the standard; however, PG&E is in compliance with the intent of the standard.

Water-Cooled Nuclear Power Plants

L-4 Attachment 2.1 Rev. 31 Page 46 of 52 Reg. D.ate Standard. Rev. Title/Subj~ct ~xceptions

~uJCI:e:S :No.

• 1 ..58 :9/80 .ANSI N45*:2~6 1978 Qualification of ANSI N45~ 2. 6 applies to individuals conducting independent QC inspections, Nuclear Power Plant examinations, and tests .. ANSI/.ANS 3.1-1978 applies to personnel conducting Inspection, inspections and tests of Items or activities for .which they are responsible (e.g.,

Examination and surveillance tests, maintenance tests, etc.).

Testing Personnel I

• Except that inspector/examiner reevaluation due dates may be extended a I maximum of 90 days. The next reevaluation due date shall be based on the original scheduled due date but shall not exceed the original due dat~ plus 90 days. I NDE personnel sha*ll be qualified and certified in accordance with CP-189-1.995. I NDE.personnel who perform examinations of the containment structure per the requi'rements of Section XI, Subsections IWE and IWL, visual examination and I ultrasonic thickness measurement only. shall be qualified and certified to ANSI/ASNT CP:..189-1991. I lSI ultrasonic examiners shall meet the additional requirements of ASME Section I

XI Appendix VIII, 2001 Edition with no Addenda. I

1'.88 10/76 ANSI N45.1.9 19.74* Collectioni Storage; For the HS ISFSI, PG&E will. comply with the 2:-hour rating of Section 5.6 of ANSI and Maintenance of N45.2.9 issued July 15, 1979. For SAFSTOR, PG&E shall comply with a 1-hour NuClear. Power Plant fire rating rather than that spedfied therein.

QualitY Assurance Records Except PG&E 'Yill also meet the intent of the guidelines for the storage of QA records in e_lectronic media as, endorsed by Generic Letter .88-18, "Plant Record Storage on Optical Disks," issued October20, 1988,. and Regulatory Issues Summary 2000-18, "Guidance on Managing Quali~y Assurance Records in Electronic Media," issued October 23, 2000 ..

Note: PG&E will maintain records of spent fuel and high-level radioactive waste in storage in accordance with ANSI N 45.2.9-1974 rather than* 10 CFR 72.72(d}.

""- -* Refer to HB ISFSl FSAR Update, Section 9.4.2.

L-4 Attachment 2.1 Rev. 31 Page 47 of 52 Reg. .Oat~ Standard :Rev. Titi~/Subject Exceptions Guides N.o.

1J'4' 2F1.4. ANS1 t973 Quality Assurance N45,2::10 Term~-' and I Definitions 1:?41 6176 ANS:I 1974 _O.uality Assurance Except PG&E will allow the *design_er~s immediat~ supervisor to. perform design N4s.z:.-1.1 Requiremf?nts. for verificatjoiJ .in exceptiomil circumstances and with the controls as described i_n the De.sJgn .of NUREG-OSOO, Revision 2, July 1981.

Nuclear Power Plants 1.144 117:9 .ANSI 1977 .AudUihg of Quality Except the scheduled date for triennial vendor audits and annual supplier N45.2.12 Assuraric.e eValuations tn~y be extended a maximum of 90 days. The next scheduled due Programs for date shall ber based .on the original scheduled .due date but shall not exceed the Nuclear Power ori§inal due date plus 90 days.

Plants Except that the corrective action pro*gram stipulated in the QA Program may be us:ed irts_tead of the requirements* o.fSection 4.5.1 as long as the appropriate time limits are applied to significant conditions adverse to quality. Also, no additional documentation is necessary if needed corrective actions are taken and verified priorto audit report issuance.

S.ee Note for Reg Guide 1.144 1..123 7177 .ANSI 1-976 Quality Assurance In *addition to ANSI. N45:2.1-3, Section 10.3.3, PG&:E w.ill accept items and services N~5.2.13 R'equirements for which are *complex or involve special processes, environmental qualification, or Control* of critical characteristics which are difficultto verify upon receipt by suppliers'

• Procurement of Certificate .of Conformance if and only if the supplier has- been evaluated and Items and Services qu~lified utilizing Performance Based Supplier Audit techniques.

for Nuclear Power*

p*rants Se:e_ Note for Reg ~uide t._1_~

L-4 Attachment 2.1 Rev. 31 Page 48 of 52 Reg-. I Date :standard Rev. Title/SubJect Exceptions Guldes No.

• 1.146 I 8/80 ANSI 1978 Qualification of Except. that auditor recertification due dates may be extended *a maximum of 90 N45.2.23 Quality Assurance days. The next recertification due date shall be based on the original scheduled Program Audit due date but shall not exceed the original due date plus 90 days.

Personnel for Nuclear Power Plants Except that in lieu of the requirements of 2. 3.4 of ANSI N45.2-1978, the prospective lead auditor sh.all have participated in at least one nuclear quality assurance audit within the year preceding the individual's 'effective date of

~ualification.

1.3*3 2/78: I ANSI N 18.7 1976 I Quality Assurance Except that PG&E will not perform biennial review of all HB ISFSI procedures, Progr.am except under the conditions described _in note below (See note at end of table).

-Requirements

_:{QperatioiJl Except that audit frequencies specified in Regulatory Guide 1.33, Revision 2, need not be met. Audits shall be performed at the frequencies specified in S.ection 18.0 of this. QA Program.

• Except-that QV audits and independent reviews .of the Emergency Preparedness Prog_mm $ball b~performed in accordance with 10 CFR 50.54ffi.

Except that a grace perio.d of up to 90 days will be allowed for .audit scheduling, except where the schedule is mandateti by regulation. The next schedule due date shall be based on the original scheduled date but shall not exceed the original due date plus 90 days.

Exceptthatwhen purchasing commercial-grade calibration services from certain accre.dited calibration laboratories, the procurement documents are not required to impose a quality assurance program consistenfwith AN.SI N45.2-1971. Alternative requirements described in this QA Program for Regulatory Guide 1.123 may be impiemented in*lie.u of imposing aqualityassurance program consistent with ANSI N45.2~1971.

L-4 Attachment 2.1 Rev. 31 Page 49 of 52 Reg. Date Standard *Rev. Title/Subject Exceptions

.G.Uides. No *

• • - NCIG*01 2 Visual Weld.

Acceptance Criteria for Structural Welding at Nuclear Power Plants

- -- NCIG-02* 2 Sampling Plan for Visual Reinspection of Welds

- - NCIG-03 1 Training Manual for Inspection- of Structural Weld at Nuclear Power Plants Using the Acceptance Criteria of NCIG.;01

L-4 Attachment 2.1 Rev. 31 Page 50 of 52 Notes for. Reg. Guide 1.33:

. Th~s~ control$ r~pl~ce. foe biennial procedur~ :review *requirem~nt found ill Section 5.2.15* of ANSI N18,7-1*976.

1. All *applicable .H B ISFSI procedures (shall)"" be reviewed following an unusual incident, such as ari accident, unexpected transient, significant operator error, or ecfuipJn.entmalfu*nctio*n, and foUowing any modification to a system, as specified by Section 5.2 ofANSI N18.7/ANS 3.2, which is endorsed by Regulatory G.uide 1.33.

2. Non-routine procedures. (e.g. emergency. operating pmqedures, procedures which implement the emergency plan, and other procedures. whose usage may be dictated by an ~vent) (shall),.. be revrewed at least every two years* .and revised as* appropriate.

3. Routine HB-ISF$1 pro~;edures fhat have not been used for two years (shall)* be reviewed before use to determine if changes are necessary or desirable ..

~The. word .sho!..lld has been;chang~d to shall denoting a regulatory commitment.

Notes, for Reg. Guide 1..1-44:

The. following interpretation is added With respectto Regulatory Gulde 1.144, Section C.3.b(2):

\1\Jiieh p*ur.chas:ihg conir:hetcial-grade*calibratlon .servi.ces from calibration laboratories accredited by a nationally-recognized body, the accreditation process and acGrecHting body *may be credited With carr)ilhg out .a portio*n of the Purchasers duties-of verifying acceptability and effective implementation of the calibration service s.uP,pliet'.s qua:lity assurance progra*m.

NationalJy.,recognized .accrediting bodies 'include the National Voluntary Laboratory Accreditation Program (NVLAP) administered by the National Institute of Standards and Technol~gy '(NIST) .and .other accrediting bodies recognized by NVLAP via a Mutual Recognition Agreement (MRA)

In lietJ of pe*rforming . an audit, acc*epting an -audit by another licensee, or performing a commercial,..gtade supplier survey, a :documented review of the suppliers' accreditation s.ha:U be *performed by the .Purchaser. This review shall jnclude, at. a minimum, verffication of all the following:

(1') Th.e ~ccr~ditatian* isJo ANSJ/ISO!IEC 17025.

• (:2)

• The. accre¢!i.tin:g bqdy is* either NVLAP or A2LA; which is an accrediting body re.Gognized by NVLAP through an MRA.

(3) The published scope .of accreditation for the calibration laboratory ,covers the needed measurement parameters, ranges, and uncertainties.

L-4 Attachment 2.1 Rev. 31 Page 51 of 52 Notes for Reg. Guide *1.123:

The requirements ofAN:Sl N452.13, Section 3.2, .,Content of the Procu*renient Documents,~* Subsection 3.2.3, "Quality Assurance Program Requirements" are accepted

.with,:the fa.llowing exception:

• When P.trrch<:lsiog *commerdaJ..;g.rade services ft*om calibration laboratories accredited. by a nationally-recognized accrediting body, the procurement documents are not re:quiredJd*lmpose. ;a ,qu.ality* ass:urance program c:onsistent with ANSI N4~.z:..te71. Nationally-recognized accrediting bodies include the NVLAP administered by the NJST' and .o.ther accredltihg bodies tecog.nized by NVLAP v.iaa MRA In* such cases, accreditation may be accepted in lieu of the purchaser imposing a QA Program corrsi.~tent .w.~th ANSI N45.Z.,.1971, provid:e.d ,aJJ tne. folloWil!g are met:

(1.) The .accre.d.itation is to .ANSI/ISO/I EC 17025 (2) The =accrediting body is either NVLAP or A2l_A, which is an accrediting body recognized by NV LAP through an MRA

• (3.) The published scope of accreditation *for the calibration laboratory covers the needed meas,urement parameters, ranges, and uncertainties~

'(4) The purchase ;documents impose *additional.techn'ical*and administrative requirements, as necessary, to satisfy HBQAP and technical requirements, including the requir~.men.'f that the ~aHbration/certificpte report include identification of the laboratory equipment/standard used .

.(5:) The :pu.rchase d.o9uments r~quire reporting as:.fourid calibration data when calibrated items are foun*d to be out-of-tolerance.

L-4 Attachment 2.1 Rev. 31 Page 52 of 52 r, '

i ~- ..:;_

App~ndix D Organization Chart- Generic Titles Board of Directors-PG&E Corporation President PG&E I

Vice President - Senior Vice President I

Vice President -

Support Services - Energy Supply Engineering I I Support Manager I Services Nuclear* Director Applied Supervisor Supply Technology Services*

Records Chain Senior Vice President

-Chief Nuclear I

I I

I Officer I I

I I I I I I I I _____________________________ J I L--------------~-------~-----------

I I Director - Quality Director-- Humboldt Nutlear SQf~ty Verification Bay* Oversight Committee (NS.OC)

I Ehgineering*

I "ISFSI Manager* Radiatioh Protection I

Independent Safety Manag~r* .Manager* Review (ISH)*

• Denotes Humboldt Bay site position