ML20045B905

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Safety Evaluation Re Order Approving Decommissioning Plan & Authorizing Decommissioning of Rsngs,Unit 1,SMUD.Concludes That Reasonable Assurance That Health & Safety of Public Will Not Be Endangered by Decommissioning Option,Provided
ML20045B905
Person / Time
Site: Rancho Seco
Issue date: 06/16/1993
From:
NRC
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Shared Package
ML20045B878 List:
References
NUDOCS 9306210244
Download: ML20045B905 (54)


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SAFETY EVALVATION BY THE NUCLEAR REGULATORY COMMISSION RELATED T0 THE ORDER APPROVING THE DECOMMISSIONING PLAN AND AUTHORIZING DECOMMISSIONING OF RANCHO SECO NUCLEAR GENERATING STATION. UNIT 1 SACRAMENTO MUNICIPAL UTILITY DISTRICT DOCKET NO 50-312 h

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SAFETY EVALUATION i

TABLE OF CONTENTS

1.0 INTRODUCTION

1

1.1 Background

1 i

1.2 Proposed Action..........................I

?O OrcrplPTTON OF DECOMMISSIONING ACTIVITIES AND TASKS 3

2.1 Decommissioning.

........................3' 2.2 Decommissioning Activities and Tasks 3

2.2.1 Custodial-SAFSTOR....................

4 2.2.2 Hardened-SAFSTOR.....................

6 2.3 Schedule............................

9 2.4 Decommissioning Organization and Responsibilities........

9-i 2.5 Training Program 10 2.6 Contractor Assistance......................

11 3.0 OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 12 3.1 Facility Radiological Status 12 3.1.1 Facility Operating History................

12 3.1.2 Radioactivity Inventories 14 3.1.3 Plant Radiation and Contamination Levels.........

17 3.1.3.1 Reactor Building Contamination Levels and Exposure Rates 17 3.1.3.2 Auxiliary Building Contamination Levels and Exposure Rates 18 3.1.3.3 Fuel Storage Building..............

20 3.1.3.4 Turbine Building 20 3.1.3.5 Areas Outside of Major Structures........

21 3.1.3.6 Inaccessible Structures / Systems.........

21 3.1.3.7 Plant Radiation and Contamination Levels Summary 22 3.2 Radiological Protection.....................

23 4.0 RADI0 ACTIVE WASTE MANAGEMENT.....................

25 4.1 Liquid Radioactive Waste 25

.l 4.2 Gaseous Radioactive Waste.....................

26

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4.3 Solid Radioactive Waste......

.............. 27 4.4 Process and Effluent Radiological Monitoring Systems 29 4.5 Spent Fuel Disposition 30 4.6 Waste Handling and Packaging 30 4.7 Waste Transportation 30 5.0 FINAL RADIATION SURVEY PLAN 30:

6.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS IN PLACE DURING DECOMMISSIONING 31 t

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'l SAFETY EVALUATION TABLE OF CONTENTS (CONTINUED) i i

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7.0 QUALITY ASSURANCE (QA)........................ 31-7.1 QA Organization.........................

31 7.2 Q A Pl a n............................. 32 8.0 POSTULATED ACCIDENTS.........................

32 9.0 FINANCIAL ASSURANCE

......................... 35.

9.' 1 Cost Estimates for Decommissioning 35 9.2 Method for Funding Decommissioning 37 9.2.1 Financial Assurance Plan................

38 9.2.1.1 Safety Considerations..............

38.

9.2.1.2 Financial Considerations

............ 39 9.2.2 Public Comments 42 9.2.3 Decommissioning Funding Summary 44

.10.0 STATUS OF DECOMMISSIONING PLAN LITIGATION.............. 44

11.0 CONCLUSION

S.............................

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12.0 REFERENCES

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h LIST OF TABLES 1.

Plant Radioactivity Inventory 14 2.

Radionuclide Inventory in Concrete and Sludge in the Retention Basin 15 3.

Radionuclide Concentrations in Soil Remaining in the Tank Farm Area..

15 4.

Radionuclide Concentrations in Clay Creek Sediment 16 5.

Reactor Building Contamination Levels.................

18 6.

Auxiliary Building Contamination Levels and Exposure Rates 19 7.

Exposure Rates for Inaccessible Structures / Systems

~21 8.

Total Estimated Volume of Solid Radioactive Waste Generated During SAFSTOR 28 9.

Total Estimated Volume of Solid Radioactive Waste I

Generated During Deferred-DECON....................

28 i

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1.0 INTRODUCTION

i 1.1 Backaround The Rancho Seco. Nuclear Generating Station (RSNGS) is located in southern Sacramento County, California.

RSNGS is described in the RSNGS Unit 1 " Updated Safety Analysis Report" (USAR) (Ref.1], and in the U.S. Nuclear Regulatory Commission staff (formerly the U.S. Atomic Energy Commission) " Safety Evaluation by the Directorate of Licensing, U.S. Atomic Energy Commission, in the Matter of Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station, Unit No.

1," Docket No. 50-312 [Ref. 2].

Significant radioactive contamination is contained in the following major structures:

(a) the reactor building, which contains the pressurized-water reactor (PWR) designed by Babcock and Wilcox; (b) the auxiliary building, which houses the chemical and volume-control system, i

the waste-handling system, auxiliary coolant systems, new fuel storage facilities, and components of the engineered safety features'; and (c) the fuel storage building which contains the spent fuel pool (SFP).

The fuel storage building is structurally separated from the auxiliary building.

Bellows are provided in the fuel transfer system, to allow for differential movement between the reactor building and the SFP, to accommodate thermal loads and loads caused by earthquakes [Ref. 2].

RSNGS operated for approximately 15 years with about 2,149 effective full-power i

days of operation, and was shut down on June 7, 1989 (" Rancho Seco Nuclear Generating Station Proposed Decommissioning Plan" [PDP]) [Ref. 3].

The Sacramento Municipal Utility District (SMUD), owner and licensee, decided to permanently shut down RSNGS.

The decision to permanently shut down the plant was based on the outcome of a June 6, 1989 referendum, and licensee inability to sell the plant. On August 29, 1989, the licensee formally informed the U.S. Nuclear Regulatory Commission (NRC) of its decision to permanently shut down RSNGS.

Because the decision to shut the plant down before the expiration of its J

operating license had not been anticipated, the licensee was unable to give NRC a preliminary decommissioning plan in accordance with 10 CFR 50.75.

The licensee submitted its " Plan for Ultimate Disposition of the Facility" (PUDF) in July 1990, in response to an NRC staff request [Ref. 4].

i The licensee proposes to decommission RSNGS using the SAFSTOR alternative. The term "SAFSTOR," in this report is intended to be inclusive of both the Custodial-SAFSTOR and Hardened-SAFSTOR subcategories of the SAFSTOR alternative, as generally defined in the " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," NUREG-0586, August 1988 [Ref. 5]. The licensee contemplates using two subcategories of SAFSTOR, that is, Custodial-SAFSTOR and Hardened-SAFSTOR, followed by Deferred-DECON.

References 3 and 6-10 l

are the basis for this safety evaluation report (SER).

1.2 Proposed Action The licensee proposed decommissioning RSNGS employing two subcategories of the SAFSTOR alternative:

Custodial-SAFSTOR and Hardened-SAFSTOR, followed by Deferred-DECON at the end of the Hardened-SAFSTOR period (approximately 20 years after final shutdown). The licensee described its decommissioning plan in its PDP [Ref. 3], transmitted by letter (D. R. Keuter, SMUD, to S. Weiss, U.S. NRC, May 20, 1991 [Ref. 11].

1 RSNGS operated for approximately 15 years and a number of areas on the plant site, in the immediate plant environs, and in plant buildings are contaminated.

Contamination is present in the plant environs, in part, because of steam generator tube leaks, a history of fuel failures, and subsequent radioactive liquid releases to the environment [Ref. 9].

Steam generator tube leaks and failed fuel also contaminated secondary and auxiliary systems in the plant. While the plant is being prepared for Hardened-SAFSTOR, the licensee will perform only decontamination activities necessary to reduce loose surface contamination to levels that will allow personnel access to buildings for routine surveillanct and maintenance. Access to contaminated areas will be controlled, and contaminated areas will be entered only for the purpose of inspection and maintenance during the SAFSTOR period [Ref. 3].

The SAFSTOR alternative selected by SMUD will culminate with the implementation of Deferred-DECON, approximately 20 years after plant shutdown.

During the Deferred-DECON phase, radioactive materials will be reduced at the site to acceptable levels, so that the license can be terminated.

The Deferred-DECON alternative is scheduled to begin in the year 2008.

The licensee will initially place the plant into a Custodial-SAFSTOR condition.

The Custodial-SAFSTOR phase will last until all spent fuel is removed from the SFP and all 10 CFR Part 72 independent spent fuel storage installation (ISFSI) issues are resolved.

Once all spent fuel has been removed from the SFP, SMUD will begin the process of placing the plant into a Hardened-SAFSTOR condition.

In preparation for Hardened-SAFSTOR, systems no longer needed to support SFP operations will be drained, de-energized, and secured.

During the Hardened-SAFSTOR period, the plant ventilation system, high-efficiency particulate air j

(HEPA) filters, and radiation monitoring system will be maintained and operated continuously.

Although the licensee analysis assumes NRC approval of an ISFSI design, this safety evaluation does not depend on that approval.

Even if RSNGS were to remain in Custodial-SAFSTOR until Deferred-DECON, the impacts of RSNGS in Custodial-SAFSTOR are well within the impacts evaluated in the Final Environmental Statement related to the operation of RSNGS [Ref. 40).

For this scenario, this Safety Evaluation herein of the Custodial-SAFSTOR would apply to the entire SAFSTOR period.

R Generally, spent fuel issues are not considered to be a part of a licensee decommissioning effort.

SMUD has chosen to make fuel disposition an integral part of its decommissioning plan.

The licensee proposes to maintain the plant in Custodial-SAFSOR while spent fuel is stored in the SFP.

When spent fuel is removed from the SFP to an ISFSI (SMUD anticipates spent fuel transfer to an ISFSI in 1998), the licensee will begin placing the plant into the Hardened-SAFSTOR phase.

The decommissioning activities contemplated by SMUD are consistent with the decommissioning alternatives defined in NUREG-0586 [Ref. 5]

and in " Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station," NUREG/CR-0130, Volume 1, June 1978 [Ref.12].

However, the transition from one SAFSTOR phase to another will depend on the licensee ability to obtain a license from the NRC pursuant to 10 CFR Part 72 for the proposed ISFSI for storage of spent fuel.

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t 2.0 DFSCRIPTION OF DECOMMISSIONING ACTIVITIES AND TASKS 2.1 Decommissioning SMUD selected the SAFSTOR decommissioning alternative for RSNGS.

SMUD has proposed the use of two subcategories of SAFSTOR; that is, Custodial-SAFSTOR and Hardened-SAFSTOR. The licensee expects that its selected SAFSTOR darnmmiscinnino alternative will end approximately 20 years after final plant shutdown [Ref. 3].

The licensee schedule is based on the assumption that all Part 72 ISFSI licensing issues will be resolved without delay. The licensee is proposing a Custodial-SAFSTOR period that will last until all spent fuel is removed from the SFP. The licensee schedule anticipated that spent fuel will be transferred to an ISFSI in 1998.

This Safety Evaluation is not dependent on the licensee being granted approval for an ISFSI.

The licensee long-range decommissioning plan envisions the construction of a dry cask ISFSI, for the storage of RSNGS spent fuel. The SMUD schedule calls for all spent fuel to be in casks by 1998. The licensee is purchasing a transportable storage cask system that will consist of two dual-purpose casks, and a sufficient number of multielement sealed canisters (MESCs) and concrete modules to store all RSNGS spent fuel assemblies [Ref. 7].

The licensee plans to store RSNGS spent fuel in a transportable MESC system differs from the concept of storing its spent fuel in a fleet of dual-purpose casks, as described in the licensee PDP [Ref. 3]. The details of spent fuel disposal are beyond the scope of the RSNGS decommissioning.

However, because solving the problem of spent fuel disposal raises issues that have the potential i

for impacting the licensee decommissioning schedule, the staff considered the potential impacts of delays on the proposed decommissioning of RSNGS.

While spent fuel is stored in the SFP, required maintenance will be performed on systems / structures needed to support SFP cooling, security, building services, environmental and radiological monitoring, and fire protection.

Systems that may be used during Custodial-SAFSTOR, Hardened-SAFSTOR, or Deferred-DECON will be maintained in a preserved condition by the licensee.

The licensee selected decommissioning alternative includes steps to remove radioactive fluids from systems, sluicing resins from tanks, removal of radioactive waste from the site, radiation and environmental monitoring, and f acility security during the entire SAFSTOR period.

The licensee commitment to the above is consistent with the guidance given in " Termination of Operating License for Nuclear Reactors," Regulatory Guide 1.86, Sections C.2.b and C.3.a

[Ref.13], and is acceptable. The licensee proposed SAFSTOR alternative will delay major dismantlement activities (DECON-decommissioning at RSNGS is anticipated to begin approximately 20 years after plant shutdown).

However, the delay is less than 60 years, and is acceptable, in accordance with 10 CFR 50.82(b)(1)(i).

2.2 Decommissioning Activities and Tasks The activities and tasks associated with the decommissioning of RSNGS involve the preparation and implementation of plans and procedures necessary to prepare

. the plant for each phase of the proposed SAFSTOR decommissioning alternative.

Preparation and planning to place the plant safely into the selected SAFSTOR conditions have been divided by the licensee.into the following four activities:

(a) preparation for Custodial-SAFSTOR, (b) Custodial-SAFSTOR, (c) preparation for Hardened-SAFSTOR, and (d) Hardened-SAFSTOR.

The licensee anticipates that the SAFSTOR period will last approximately 20 years.

At the end of the SAFSTOR period, the licensee will enter the Deferred-DECON period during which the licensee plans to commence reducing radioactive materials at the site to acceptable levels and preparing for license termination.

The licensee will submit an updated decommissioning plan that will describe the activities during Deferred-DECON.

A generic determination associated with temporary storage of spent fuel after the cessation of reactor operation [see 10 CFR 51.23(a)] has been made by the l

Commission.

The Commission has determined that spent fuel can be safely stored ~

in the reactor plant SFP, or in an ISFSI at the reactor site, or in an ISFSI away from the reactor site, for up to 30 years after the end of reactor operations.

Based on this Commission generic determination, the staff concludes that the fuel storage methods proposed by SMUD are safe and environmentally acceptable.

SMUD committed, in its " Financial Assurance Funding Plan," to review decommissioning costs annually, and to revise the annual contribution to the decommissioning trust fund every 5 years [Ref. 7]; the staff considers this approach acceptable and in accordance with 10 CFR 50.82(c)(2).

The activities and tasks associated with each phase of the licensee selected SAFSTOR alternative were evaluated.

Detailed evaluations of Deferred-DECON activities and tasks are outside the scope of the SAFSTOR decommissioning alternative selected for RSNGS. The details necessary for the staff to evaluate Deferred-DECON will be provided when the licensee submits an updated decommissioning plan.

This is acceptable in accordance with 10 CFR 50.82(d).

2.2.1 Custodial-SAFSTOR Decontamination during the preparation for Custodial-SAFSTOR will be limited to that necessary to maintain exposures as low as is reasonably achievable (ALARA),

and to clean surface contamination from building access pathways, to enable

'J entry into areas within the radiologically controlled areas (RCAs), without the need for-donning anticontamination clothing.

Even before the decommissioning order is issued, decontamination is being conducted, to the extent necessary, to=

. reduce the potential for migration of contamination.

Throughout the entire i

SAFSTOR period, plant decontamination efforts generally will be limited to simple decontamination techniques of vacuuming, mopping, or scrubbing with I

cleaning agents compatible with the waste treatment system.

i However, mechanical decontamination of external surfaces'will be conducted, where necessary, on surfaces that contribute significantly to radiation exposure to surveillance and maintenance personnel, during the Custodial-SAFSTOR period

[Ref. 3].

The licensee will maintain sufficient onsite staff,-during the Custodial-SAFSTOR period, to perform radiological surveillance and any other

. decontamination practices necessary to maintain plant radiological conditions.

During the Custodial-SAFSTOR period, the licensee will maintain -all systems and components required to support spent fuel storage in the SFP, in accordance with Amendment No. 119 to Facility Operating License No. DPR-54, Rancho Seco Nuclear Generating Station, " Permanently Defueled Technical Specifications," March 19, 1992 [Ref. 14].

The following systems will be drained, flushed, and left in place during the Custodial-SAFSTOR period:

reactor coolant system (RCS) safety injection and makeup (SIM) system core flood system (CFS) decay heat system (DHS) purification and letdown system (PLS) pressurizer relief tank (PRT) reactor coolant drain (RCD) system containment building spray (CBS) system The auxiliary boiler will be maintained as long as it is needed for processing radioactive waste.

Propane will be disposed of, liquids will be drained from systems, and hydrogen will be removed from the hydrogen gas system to prepare for Custodial-SAFSTOR. The licensee will vent the nitrogen gas system, and all systems using nitrogen will be isolated during the Custodial-SAFSTOR phase of decommissioning [Ref. 3].

The licensee proposes to keep the drain and sewage, miscellaneous liquid radwaste, blender / dryer, and solid waste systems functional during Custodial-SAFSTOR.

The waste gas system will be taken out of service, and waste gas surge and decay tanks will be vented and isolated [Ref. 3].

The licensee will keep the reactor and auxiliary building HEPA filters functional, and continue to perform in situ testing and periodic maintenance, during Custodial-SAFSTOR. The turbine building exhaust fans, supply fans, and air handlers for the emergency pump room cooler will be deenergized [Ref. 3].

The licensee proposes to preserve needed portions of the liquid radwaste system, and the HEPA filters for the reactor and auxiliary buildings.

Further, the I

licensee proposes to maintain all systems and components required to keep the SFP in accordance with approved technical specifications (see Amendment No.119

[Ref. 14]), and to retain sufficient onsite staff to continue surveillance and maintain security.

During the Custodial-SAFSTOR period, cranes will be required for moving heavy components and be maintained for that period. The licensee will maintain the.

cranes in the fuel storage building, as long as there is fuel in the SFP. After the spent fuel is removed from the SFP, the SFP will be drained, the water will i

be processed, and the fuel-handling systems will be left in place [Ref. 3].

The licensee will empty and deenergize the main carbon dioxide (fire suppression) system. The carbon dioxide and halogen subsystem in the record storage vault will be maintained as long as records are kept there. The water

. i supply, pumps, hydrants, and underground mains will be maintained.

Electrical systems not supporting functional systems will be deactivated and abandoned in place during Custodial-SAFSTOR. The plant communication system will be main-tained in a modified configuration, to support the emergency and security plans

[Ref. 3].

The licensee estimates that it will require approximately 427,100 manhours and approximately 31.6 man-rem to place RSNGS in Custodial-SAFSTOR [Ref. 36]. The exposures during Custodial-SAFSTOR can be broken down as follows:

16.6 man-rem during preparation for Custodial-SAFSTOR, and 15.0 man-rem for the dormancy phase of Custodial-SAFSTOR.

During the Custodial-SAFSTOR period, the licensee will maintain full-time onsite surveillance by operating and security personnel.

The licensee will continue to monitor radiation and maintain required equipment, and will take measures to prevent accidental or deliberate intrusion [Ref. 3].

The licensee estimates that plant staff would receive a significant radiation dose if the reactor vessel were completely drained. The licensee intends, during the SAFSTOR period, to fill the reactor vessel with water to a level up to approximately the decay heat line [Ref 3]. Water remaining in the reactor vessel will be at a level below the " hot leg" nozzles, and this volume has been calculated to be approximately 22,000 gallons.

Evaporation is a credible way for water to leave the reactor vessel. Water vapors from the reactor vessel will migrate from the reactor vessel into the steam generator, condense, and gravity-drain through the normal reactor coolant system (RCS) drains into the radioactive waste system [Ref. 8].

The cooling towers at RSNGS contain sheets of Transite, a material commonly called " fill," which contains a significant quantity of asbestos.

During preparation for Custodial-SAFSTOR, the fill material will be removed for disposal by certified contractors, registered by the State of California. The cooling tower basins will be kept filled with water, to provide a source of water for fire protection during the SAFSTOR period [Ref. 3].

SMUD decommissioning commitments related to activities and tasks during the Custodial-SAFSTOR phase were compared to decommissioning elements in Table 2.4-1 of NUREG-0586, and the description of Custodial-SAFSTOR in NUREG-0586 [Ref. 5].

The licensee committed to maintaining key elements of the plant, as defined in NUREG-0586, such as: maintaining full-time onsite staff for surveillance; using operating and maintenance staff; maintaining the ventilation system operable; and constructing barriers that would prevent accidental or deliberate intrusion.

On the basis of its review and this comparison, the staff finds that the licensee approach to key Custodial-SAFSTOR activities and tasks is acceptable.

2.2.2 Hardened-SAFSTOR The licensee plans to maintain areas, structures, and systems within the industrial area for restricted use, during Hardened-SAFSTOR. These areas, structures, and systems will be decontaminated, as necessary, to reduce, or stabilize significant radioactive contamination, to ensure the safety of the public and workers [Ref. 7].

. Doring the preparation for Hardened-SAFSTOR, the licensee will implement plans.

tc harden the reactor building, fuel storage building, and auxiliary building.

The licensee will perform the following activities [Ref. 3]:

Doors to noncontaminated areas that do not require routine access will be locked and secured.

Doors to areas containing radioactive materials or other contamination will hn secured, to prevent accidental intrusion and to make deliberate intrusion very difficult.

The contamination levels of the turbine building and the interim onsite storage building (IOSB) will be reduced and the buildings will be secured.

Shielding will be added where necessary, to maintain radiation exposures to security guards and plant personnel ALARA.

When the need for radwaste processing is reduced, the small auxiliary boiler will be drained and deenergized, and both boilers will be left in place.

In addition, power supplies to secondary side pump motors will be isolated and the main condensate and makeup system components will be deenergized and isolated.

While preparing for Hardened-SAFSTOR, the SFP will be drained, and the contaminated walls, grates, and associated components will be cleaned or stabilized, to prevent the spread of contamination.

The SFP filter demineralizer resins will be removed and the entire system will be isolated after processing SFP water through the radwaste system [Ref. 3].

The major activities necessary to place RSNGS into Hardened-SAFSTOR can be summarized as follows [Ref. 6]:

Transfer spent fuel from the SFP to dry-cask onsite storage.

Drain and process the water from the SFP.

Drain, deenergize, and secure all systems not needed to support the security and surveillance program.

Perform a radiation survey of the plant.

l Lock buildings containing radioactive materials (e.g., reactor, auxiliary, and fuel storage buildings), to avoid accidental intrusion.

During Hardened-SAFSTOR, radioactive waste equipment, tanks, and piping will be drained.

Ion-exchange resins will be sluiced from vessels, all water processed, and electrical circuits deenergized.

The waste gas system will be left in place with no maintenance, and the sewage and drain systea will be maintained, as needed. The licensee will keep the radiation monitoring system functional during Hardened-SAFSTOR.

The 480 volt power supply system will be kept functional, and the plant communication system will be maintained in a configuration that will support the emergency and security plan [Ref. 3].

SMUD has defined the Hardened-SAFSTOR that it intends to implement at RSNGS to be a more active form of Hardened-SAFSTOR than the Hardened-SAFSTOR defined in NUREG-0586 [Ref. 5] and NUREG/CR-0130 [Ref. 12]. SMUD has committed to maintaining the following ongoing activities while in Hardened-SAFSTOR [Ref. 7]:

Continue to monitor the ventilation exhaust from significantly contaminated bt.il d i ng s.

Perform preventive and corrective maintenance on required security systems, area lighting, and general use buildings.

. Maintain a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> staff.

Continue routine radiological inspections of contaminated buildings.

Maintain the structural integrity of buildings.

SMUD has committed to continue to maintain and operate the ventilation exhaust systems in buildings containing significant' amounts of radioactivity [Ref. 7].

SMUD has also committed to maintaining the radiation monitors associated with the ventilation exhaust system, to provide assurance that plant radioactive release points are monitored.

In addition, licensee personnel will continue surveillance of inactive systems and structures, to ensure system and structural integrity, and to provide assurance that radioactivity does not migrate from contaminated systems and structures during the Hardened-SAFSTOR period [Ref. 7].

The licensee has also committed to maintain the following site programs in effect during the Hardened-SAFSTOR period [Ref. 7]:

(a) radiation protection program (RPP), including the radiological environmental monitoring program (REMP); (b) Technical Specifications surveillance program; (c) fire protection program; (d) security prograa; (e) quality assurance program; and (f) training program, including certified fuel handlers and general employee training.

The guidance in NUREG/CR-0130 [Ref. P recommends that, during Hardened-SAFSTOR, the licensee extensively dec s taminates and cleans up areas containing i

significant quantities of radioactivity.

SMUD has committed to perform some decontamination, erect barriers, and seal in some contaminated areas durir.g the Hardened-SAFSTOR phase.

To compensate for the limited decontamination at RSNGS during Hardened-SAFSTOR, SMUD has committed to maintain sufficient staff on site to continue maintaining systems and to provide radiological surveillance, to ensure that radioactivity is not spread from sealed areas of the plant to the site or the environment [Ref. 6].

The licensee has also committed to continue to monitor the ventilation exhausts from contaminated buildings, and to maintain the structural integrity of contaminated buildings during Hardened-SAFSTOR, in.a manner consistent with the more restrictive conditions of Custodial-SAFSTOR and plant operations.

Further-more, the licensee has committed to continued use of the existing RPP, REMP, and use of the Offsite Dose Calculation Manual (0DCM), to provide assurance that workers, the public, and the environment will be protected during Hardened-SAFSTOR.

For all phases of the SAFSTOR decommissioning at RSNGS, the licensee estimates that 134 man-rem will be expended. The licensee compared the 134 man-rem to the values provided in NURE.-0586 [Ref. 5].

The 134 man-rem for the anticipated 16 years of SAFSTOR at RSNGS is less than the dose estimates in Table 4.3-2, in NUREG-0586 [Ref.5].

' Significant contamination, as used here by SMUD, is contamination that exceeds the levels in NRC Regulatory Guide 1.86, " Termination of Operating i

!4 censes for Nuclear Reactors," Sections C.2.B and 3.a [Ref. 13].

_g-The licensee made commitments, in its PDP, related to major activities and tasks, auring Custodial-SAFSTOR and Hardened-SAFSTOR, that are consistent with the guidance given in NUREG-0586 [Ref. S] and NUREG/CR-0130 [Ref. 12]. The primary differences between the licensee commitment and guidance in the references above is the licensee commitment to operate and maintain the ventilation exhaust system from the most contaminated buildings (e.g., the reactor, auxiliary, and fuel storage buildings) through all phases of the SAFSTOR period. The licensee commitments to maintain the ventilation exhaust systems and the associated HEPA filters and to continue to monitor releases are acceptable for ensuring that there are no unmonitored release points from the plant, during all phases of SAFSTOR decommissioning.

The licensee addressed all anticipated major decommissioning activities and tasks for all phases of the RSNGS SAFSTOR decommissioning (e.g., Custodial-SAFSTOR and Hardened-SAFSTOR) in its PDP [Ref. 3], and in responses to staff questions [Refs. 6, 7, and 8].

The licensee provided sufficient information in the sources listed above to allow the staff to evaluate the licensee estimated radiation exposure rates (mrem /hr) and exposure time. The licensee provided man-rem and man-hour estimates for all SAFSTOR phases.

The licensee based exposures rates, during SAFSTOR decommissioning, on actual radiation levels measured in the plant.

The licensee provided data in sufficient detail related to its estimated radiation doses during all phases of the RSNGS SAFSTOR decommissioning.

The data provided included exposure time (man-hour) and dose (man-rem).

Occupational exposures during the decommissioning of RSNGS are comparable to those in Section 4 of NUREG-0586 [Ref. 5], and its basis document NUREG/CR-0130

[Ref. 12].

On the basis of its evaluation of the licensee data, the staff concludes that the licensee decommissioning plan for major decommissioning activities and tasks is acceptable.

2.3 Schedule J

The proposed schedule for decommissioning RSNGS is in accordance with the guidance in Draft Regulatory Guide, " Standard Format and Content for Decommissioning Plan for Nuclear Reactors" [Ref.15]. The schedule, especially the transition from Custodial-SAFSTOR to Hardened-SAFSTOR, and the length of the Custodial-SAFSTOR period directly depends on the licensee ability to get the necessary NRC approvals for dry onsite storage of spent fuel. On the basis of the information submitted by the licensee, the staff finds the proposed schedule acceptable even if approval for onsite storage is not achieved before the DECON phase.

2.4 Decommissionina Oraanization and Responsibilities The licensee has proposed two organizations for SAFSTOR:

one for the Custodial-SAFSTOR, and a second for the Hardened-SAFSTOR phase of decommissioning.

The organization for Hardened-SAFSTOR reflects a somewhat smaller organization than the organization in place during Custodial-SAFSTOR. The licensee Custodial-SAFSTOR organization consists of operations and maintenance staff that will use its detailed knowledge of the facility. Contract specialists and consultants will be obtained to support the decommissioning staff.

i J During decommissioning, the General Manager is the chief executive officer for 1

the licensee. The Assistant General Manager (AGM), Nuclear is the onsite i

manager responsible for decommissioning activities. When the AGM is absent, the Plant Closure Manager (CM) is the manager with overall responsibility for i

decommissioning activities. The AGM Nuclear is responsible for implementing the licensee decommissioning plan during the preparation for Custodial-SAFSTOR.

The licensee intends to perform most of the decommissioning activities during the Custodial-SAFSTOR and Hardened-SAFSTOR period. During the period through the end of the preparation for Custodial-SAFSTOR, the CM will be responsible for plant operations equipment maintenance environmental monitoring radiation protection plant chemistry analysis technical and engineering support of nuclear, electrical, and mechanical systems site security.

e At the beginning of the Custodial-SAFSTOR, the CM will assume the AGM responsibilities.

During the Deferred-DECON, the licensee expects to use a

+

decommissioning operations contractor (DOC). The Nuclear Safety Review and Audit Committee (NSRAC) will monitor the decommissioning, to ensure that operations are performed safely.

Management personnel will meet or exceed the minimum qualifications for education, training, and experience, as defined in

" Selection and Training of Nuclear Power Plant Personnel," American National Standards Institute (ANSI) Committee N18, Design Criteria for Nuclear Power Plants, ANSI N18.1-1971 [Ref.16] for comparable positions.

NSRAC members will meet or exceed the minimum qualifications of " Selection, Qualifications, and Training of Personnel for Nuclear Power Plants," ANSI /ANS 3.1-1981, Section 4.7.2 [Ref. 17]. NSRAC will review and audit major decommissioning activities involving special nuclear materials, radioactive materials, radiological controls, procedures, records, reportable occurrences, and changes made in accordance with 10 CFR 50.59.

The staff evaluation of the licensee decommissioning organization finds that the licensee corporate and project organizations are acceptable, based on the applicable provisions of " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-0800 [Ref.18], Section 13.1.1,

" Management and Technical Support Organization," and Sections 13.1.2 and 13.1.3,

" Operating Organization."

2.5 Trainino Prooram i

The licensee will maintain the following training programs during the Custodial-SAFSTOR period:

(a) general employee training (GET), (b) radiation protection / chemistry technical training, (c) operator training, and (d) maintenance training. The training at RSNGS during Custodial-SAFSTOR will be conducted by journeyman-level instructors. The Custodial-SAFSTOR staff will primarily comprise individuals who were involved with the plant during opera-tions and individuals who were responsible for maintenance [Ref. 3].

. The licensee will provide additional training, as necessary, during Custodial-SAFSTOR in the following areas.:

first aid fire brigade and fire protection emergency plan security quality assurance radioactive waste dosimetry ALARA safety 4

=

hazardous materials handling The GET program comprises two categories--Category I, for nonradiation workers, and Category II, for radiation workers.

Individuals who routinely handle contaminated or radicar.tive materials must satisfactorily complete both 1

categories of training [Ref. 3].

Radiation protection / chemistry technicians, both licensee staff and contractor, will be ANSI-qualified. These technicians will be trained on a' continuing basis in the classrocm, on the job, and through reading assignments.

Non-certified operators will be qualified in accordance with the requirements' of the permanently defueled technical specifications (PDTS). Training will be continuous and will consist of classroom and on the job training, and reading assignments. Certified fuel handlers (CFHs) will be certified by RSNGS management and will meet the requirements of the PDTS. Training will be continuing and will consist of classroom and on the job training, and reading assignments.

The NRC staff approved the licensee CFH Training Program, to replace the 10 CFR Part 55 NRC-licensed operator program, in a letter from 1

S. W. Brown, to J. R. Shetler, dated March 19, 1992 [Ref. 19].

Personnel performing maintenance functions will be trained in areas related to hazards and safety precautions, according to the type of work they will perform.

Having compared the licensee training program to the applicable sections of NUREG-0800 [Ref. 18), Section 13.2.2, " Training for Non-Licensed Plant Staff,"

the staff concludes that the licensee training program is acceptable.

2.6 Contractor Assistance The licensee intends to perform most of the Custodial and Hardened-SAFSTOR activities with plant staff. Specialty contractors may be used to perform special services during all proposed decommissioning phases. Decommissioning contractors of the following types may be used during the decommissioning:

(a) DOCS, (b) construction services, (c) radwaste management, (d) decommis-sioning specialist, (e) radiation protection, and (f) project management i

[Ref. 3]. The specialty contractors proposed to be hired by the licensee will perform services similar to those detailed in NUREG/CR-0130 [Ref. 12].

The j

information provided by the licensee, relative ~to contractor assistance, is accept able.

1 l

3.0 OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.1 Facility Radiolooical Status Knowledge of the existing plant radiological conditions is important to ensure the safety of workers and the public during decommissioning..The goal of the SAFSTOR decommissioning alternative is to achieve a condition that ensures that the safety of the public is not endangered by the residual radioactivity left at the plant.

3.1.1 Facility Operatina History RSNGS achieved initial criticality on September 16, 1974, and began commercial operation on April 10, 1975. The plant operated for approximately 15 years, i

Over the 15 years of operation, the plant operated for approximately 2149 effective-full-power days (EFPDs), and 7 refueling cycles [Ref. 3].

t During 1975 and 1976, the plant experienced two long outages, as a result of steam turbine and main generator seal oil contamination.

The plant did not achieve full power until March 1976.

RSNGS was shut down in April 1976 to verify the integrity of surveillance specimen holder tubes in the reactor vessel. While the plant was shut down, an inspection revealed that the main generator stator was severely damaged.

Full power was not reached again until October 1976 [Ref. 3].

Failure of a site transformer in 1978 limited the capacity factor of the plant to 70 percent.

In 1980, failure of the turbine rotors, leaks in steam generator tubes, modifications to the auxiliary feedwater system, and various other problems kept the annual capacity factor of the plant to a low value. An unusual event in 1985 (the complete loss of the integrated control system) and the subsequent overcooling during shutdown, forced the plant into shutdown for 1

2 years, while the licensee conducted comprehensive system reviews and testing.

By referendum, in 1989, the SMUD ratepayers voted to discontinue operation of the RSNGS.

In August 1989, the licensee formally informed NRC of its intent to permanently shut down RSNGS and to start plans to decommission the plant. All fuel was removed from the reactor vessel and placed into the SFP [Ref. 3], and the plant was placed into protective layup in December 1989, j

During plant operation, there were fuel failures and primary-to-secondary leaks (steam generator tube failures) that led to releases of radioactive materials to the immediate environs of the plant.

Liquid and gaseous radioactive materials released from RSNGS during plant operations were within regulatory limits (10 CFR Part 20, Appendix B, Table II, Columns 1 and 2).

To support this statement, the licensee submitted a compilation of its semiannual effluent reports for the years 1974 through 1991 in " Response to Request for Additional

]

Information in Support of the Rancho Seco Decommissioning Plan and Associated Environmental Report," April 1992 [Ref. 6].

t

' la accordance with'10 CFR 20.201(b)(2), the licensee is required to make surveys that are sufficient, under the given circumstances, to evaluate the extent of 1

the radiation hazards.

In a report, " Radiological Characterization Plan for the Rancho Seco Nuclear Power Generating Station", April 1992 (Preliminary)

[Ref. 20], the licensee described a five-phase plant radiological character-ization program.

The five phases include a " Quick Look," Phases I, II, III, and Final Survey [Ref. 20].

The " Quick-Look" survey was conducted (April 4-14, 1990) in support of l

preparation of cost estimates for decommissioning RSNGS.

The " Quick-Look" was designed primarily to be an initial data-gathering effort to serve as input for the Phase I characterization.

I The Phase I characterization (April 14-May 9, 1990) was conducted to provide assurance that the health and safety of the public and RSNGS workers are protected during all SAFSTOR phases. Data gathered during the " Quick-Look" were evaluated by the licensee during Phase I characterization.

Phase I characteri-i zation involved a complete walkdown of the plant and property. The walkdown included an assessment of the physical condition of the plant and a review of all known radiologically controlled areas (RCAs).

Phase I characterization included a survey of plant structures, plant systems and components, and onsite exterior areas and systems.

The structures referenced above include the reactor building, auxiliary and fuel buildings, and the 10SB. The 10SB is for onsite interim storage of low-level radioactive waste.

Phase I radiological characterization of structures corroborated existing radiological conditions reported from data from previous surveys and filled data voids.

Phase I characterization included an evaluation of cubicles that have been previously posted as containing airborne particulate concentrations in excess of 10 CFR Part 20 maximum permissible concentrations (MPCs), or that contain evidence of ceiling leaks from the floor above.

In cubicles where ambient exposure measurements exceeded 100 mR/hr, ALARA considerations were used during the data-gathering phase of " Quick-Look."

In high-radiation areas, if ALARA dose assessment warranted, existing data were used without further charac-terization.

Phase I characterization was used to verify previous decontamination efforts, and to determine if additional contamination remained. Onsite areas (e.g.,

areas outside of major structures) of known or suspected contamination are (a) retention basins, (b) tank farm, (c) cooling towers, (d) radioactive waste storage area, (e) storm drains, (f) regenerant holdup tank "A" area, and (g) the immediate environs of the plant [Ref. 20].

Phase II characterization will be implemented to establish a radiological baseline during the SAFSTOR period (e.g., Custodial-SAFSTOR through Hardened-SAFSTOR).

Baseline surveys will be used to provide a basis for quantification of any radiological changes that may occur during the inactive phase of SAFSTOR.

. Phase III characterization will be conducted before the start of Deferred-DECON, i

in support of engineering designs necessary for health and ' safety assessments for Deferred-DECON.

A final survey will be conducted at the completion of Deferred-DECON.

In its updated decommissioning plan, the licensee.will describe its planned final radiation survey in accordance with 10 CFR 50.82(b)(3). The licensee anticipates a report compiling post-decontamination monitoring and survey results [Ref. 20].

On the basis of its evaluation of information submitted by the licensee in the references given above, the staff finds that the licensee provided sufficient information to meet the requirements of 10 CFR 50.75(g).

3.1.2 Radioactivity Inventories At the time of final shutdown, a significant quantity of radionuclides remains at a nuclear power station.

The irradiated fuel stored on site contains very large q':antities of radioactivity, some of which may be released if the fuel assemblies are damaged.

Large immobile quantities of radioactivity are contained in the neutron-activated structural materials in and around the reactor pressure vessel.

Quantities of radioactivity are contained in the corrosion film on the inside of system piping.

The licensee has also identified radioactive contamination in areas outside the major structures and buildings, and in the immediate environment of the plant in

" Supplement to Applicant's Environmental Report--Post Operating License Stage"

[Ref. 9].

Table 1 summarizes the major plant radioactive inventories at RSNGS.

Radionuclide inventories based on swipe analysis for removable contamination were not carried out at RSNGS [Ref. 3].

Table 1 Plant Radioactivity Inventory location Curies Spent fuel in SFP (493 spent [uel assemblies) 140,800,000 ReactorpressurevesselfRPV) 2,582,714 Non-fuel assemblies, SFP 97,000 3

Plant systems 4,490 Primary shield wall' 524 5

SFP walls, racks, & related equipment 47 Includes the RPV and internals. Neutron activation calculations were performed using the following computer codes ANISN W CCC-255C MICRO, *ANISN-W

  • Multi-Group One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," Dak Ridge National Laboratory Radiation Shielding Information Center, February 1986 !Ref. 213, and ORCEN2 CCC 371 MICR0 "0RIGEN2 - A Revised and Updated version of the Oak Ridge Isotopic Generation and Depletion Code," Dak Ridge National Laboratory Radiation Shielding Information Center, October 1987 IRef. 221.

2 Includes 181 orifice rod assenblies, burnable poison assenblies, in-core instruments, and retainer assenblics.

3 Based on " Residual Radionuclide Distribution and Inventory at Rancho Seco Nuclear Generating Station,"

PNL 5146 IRef. 23), and Phase 1 characterization.

Includes reinforced steel and corrugated steel liner.

5 these data were calculated using an analysis of samples from the underwater vacuum fitter and estimated crud volure.

, The licensee did not analyze, as part of its RSNGS plant characterization program, concrete in the major buildings and structures.

To indicate the radioactive contamination associated with concrete, the licensee included data from a 1983 study [Ref. 3]. The data indicated that the majority of the radioactive contamination in concrete is in the first few centimeters from the surface.

Deeper migration occurs through cracks in the concrete.

Onsito cont amination outside of the major structures (e.g., reactor, auxiliary, and fuel-handling buildings) was identified in the retention basins, tank farm, storm drains, regenerant holdup tanks (RHUTs), and radioactive waste storage area. Soil samples around the two retention basins show contamination levels that range from background up to 4.9 pCi/g.

Basin sludge and concrete samples are analyzed in Table 2.

Table 3 lists the specific activity of soil in the area of the tank farm.

Table 2 Radionuclide Inventory in Concrete and Sludge in the Retention Basins' Nuclide Curies 00-60

1. 5X 10

Cs-134 2.1X10-'

Cs-137 3.1X10'3 Total 3.4X10'*

As of July 1, 1990. Estimated voltanes of contaminated sludge and concrete are 15 f t' and 236 f t',

respectively. These data were calculated on the basis of analysis of samples collected by plant personnel.

Table 3 Radionuclide Concentrations in Soil Remainina in the Tank Farm Area' Nuclide (pCi/g)

Co-60 6.0X10

Cs-134

1. 0X 10~ '

Cs-137 2.00 An estimated 110.25 f t.' of contaminated soil was removed from the tank f arm area.

l

L

, The licensee reported finding fixed contamination in the area of the RHUT.

The contamination ranged from 100 to 400 counts per minute (cpm) above background radiation.

The activity level in the storm drains was above background.

The i

concentrations for Cs-134, Cs-137, and Co-60 in the storm drains are 0.1 pCi/gm, 0.9 pCi/gm, and 0.2 pCi/gm, respectively.

1 t

The radioactive waste storage area was used for storing packaged low-level waste for shipment off site between 1975 and 1989. The area was cleaned because of being a potential ISFSI location.

The cooling towers were sampled in May 1990, to determine if the sludge at the ~

bottom of the towers was contaminated.

Sampling results have shown that the activity levels were less than the measured environmental levels elsewhere on the site of 0.3 pCi/g of Cs-137 and 0.04 pCi/g of Cs-134 [Ref. 3].

The licensee detected radioactivity levels above background along Clay Creek.

Table 4 lists the most recent maximum radionuclide concentrations in the Clay Creek sediment.

Table 4 Radionuclide Concentrations in Clay Creek Sediment' Nuclide (pCi/gm).

Co-60 1.47 Cs-134 1.20 Cs-137 11.00 Within 0.5 kilometers of the R$NCS release point.

Th most recent detailed analysis of Clay Creek sediment was conducted by the Law.ence Livermore National Laboratories in 1989, " Environmental Radiological Studies in 1989 Near the Rancho Seco Nuclear Power Generating Station,"

November 1990 [Ref. 24], and shows that the maximum radionuclide contaminations in the creek sediment were found within 0.5 kilometers of the plant release point [Ref. 3]. The licensee used the radionuclide concentrations in Table 4 to calculate the hypothetical doses to a person off site. (dose calculated for a person standing on the sediment). The licensee calculated the 1.38 mrem /yr dose using the methodology in " Calculation of Annual Dose to Man from Routine Releases of Reactor Effluent for the Purpose of Evaluation Compliance with 10 CFR Part 50, Appendix I" [Ref. 25]. The 1.38. mrem /yr [Ref. 6] is significantly smaller than the direct radiation from the uranium fuel cycle limit of 25 mrem (total body or any organ) in a calendar year. The 25 mrem limit meets the requirements of 40 CFR Part 190 that has been incorporated into 10 CFR Part 20.

. 3.1.3 [lant Radiation and Contamination levels The structures that contain significant levels of radioactive contamination are the reactor, fuel storage buildings, and the RCAs' of the auxiliary building.

In the discussion that follows, the staff summarizes the contamination levels and 1

2 exposure rates for the contaminated areas of RSNGS.

a.2.s.2 Reactor Buildina Contamination levels and Exposure Rates IRef. 31 i

i The +6-ft elevation of the reactor building includes an area of approximately

)

2 20,000 ft.

The +60-ft elevation provides access to the D-rings, the missile j

shields, and the polar crane. On this elevation, the highest contact measurement was 1800 mR/hr.

The range of contamination level and exposure rates for this elevation are given in Table 5.

The +40-ft elevation of the reactor building provides access to the auxiliary building personnel hatch and fuel transfer canal. There are approximately

'15,500 ft of surface area. Table 5 shows the contamination levels and the measured exposure rates.

2 There are approximately 16,500 ft of surface area on the reactor building

+20-ft elevation.

The highest contact measurement was found on a seal injection valve (500 mR/hr). Table 5 lists the contamination levels and exposure rates for this elevation.

2 The grade elevation of the reactor building has approximately 16,000 ft gf surface area, and provides access to the equipment hatch, the reactor building cooling unit ducts, and core flood tanks. Table 5 provides contamination levels and exposure-rate measurements for this area.

Access to the primary and secondary manways for both steam generators is on the ft elevation of the reactor building. The elevation has approximately l

2 16,200 ft of surface area. A contact exposure rate of 4000 mR/hr was measured on the "A" steam generator head. The highest contact reading was measured on the 14-ft elevation.

The reading was 250 R/hr contact (2 inches from source),

and it was found on the pressurizer surge line drain valve.

Table 5 lists the general contamination levels for this area.

2 The distances associated with the exposure rates (mR/hr) in this SER are based on the distances in air from sources to an individual in the general areas indicated (both in the text and in the tables in this SER).

The exposures are whole body exposures.

. Table 5 Reactor Building Contamination Levels i

Contamination Exposure Rate dom /cm]

leve mR/hr Location (Room and/or Componenti

+60-ft level 1K-60K 1-12

+40-ft level Operating floor

<1K-6K 0.2-45 Components 30K

+20-ft level Floor & walls

<1K-lK 0.2-5 Pipe components IK-8K Grade level Floor

<1K-4K 0.2-2 Cable trays 4K-26K

- ft level 2K-180K 2.0-500 "A" D-ring "A" reactor coolant pump 25K-90K 1.5-40

' ft level area ' area) 2K 5.0-40

- ft level (OTSG 40K 5.0-150

+40-foot level (OTSG' area) 15K-300K 2.5-150 "B" D-ring "C" reactor coolant pump motor 60K "D" reactor coolant pump seal area 20K 1

"B" OTSG 20K-90K Pressurizer 40K-60K Floors & walls 8K-110K

)

RCS piping 10K-40K 415-ft level 10K-40K Once-through steam generator.

Hot-spotexposureratesassociatedwithtpeRCSrangedfrom15mR/hrto5R/hr.

A maximum contact exposure rate of 2.5X10 mR/hr (reading at 2 inches) was recorded at RCS 40 and RCS 41, 3.1.3.2 Auxiliary Buildina Contamination levels and Exposure Rates IRef. 31 The control room, the technical support center, the normal access point to RCAs, the radiochemistry and secondary chemistry labs, the instrument repair room, and the source room are all located on the +40-ft elevation of the auxiliary build-ing. Table 6 lists the contamination levels and the exposure rates.

Contamination and exposure rates on the floors, walls, and ceilings are low on this elevation (e.g., below the levels noted in Regulatory Guide 1.86

[Ref. 13]).

l

t The auxiliary building +20-ft elevation provides access to the reactor building ventilation system, spent fuel ventilation system, and auxiliary building system. Table 6 provides the contamination levels and the exposure rates for this area.

Area exposure rates were found to be less than 2 mR/hr.

The +10-ft elevation contains the chemical storage balcony and the access to the makeup tank filter. As a result of decontamination and filter removal from this 2

,re2, cent =~ir3 tion levels were reduced to 1000 dpm/100 cm.

Exposure rates were measured at less than 2 mR/hr.

Contamination levels and exposure rates at grade level of the auxiliary building are provided in TrSle 6 The floor area at this elevation measures j

approximately 29,260 ft. Area exposure rates in the makeup valve gallery averaged approximately 4 mR/hr, with the highest reading at 10 mR/hr.

The -20 and ft elevations of the auxiliary building contain a number of the most contaminated rooms and cubicles in the auxiliary building.

Table 6 provides a suramary of the rooms and cubicles, and their contamination levels and the measured exposure rates. A 50-R/hr contact exposure rate was found on the surface of the miscellaneous waste tank.

2 There are approximately 7900 ft of surface area on the ft elevation of the auxiliary building. The highest contact exposure rate was measured on a decay heat system valve (800 mR/hr). Table 6 provides a summary of the contamination levels and the exposure rates for this area.

Table 6 Auxiliary Buildina Contamination levels and ExDosure Rates Contamination Exposure 2

Location (Room and/or Component)

Level, dpm/cm Rate, mR/hr

+40-ft level Radiochemistry lab

<1K 0.01-0.08 Access control point

<1K 0.01 Change room

<1X 0.01 Instrument repair room

<1K 0.01-0.04 Radioactive source room

<lK 0.01-0.04

+20-ft level Ventilation equipment room

<1K-6K 0.4 Radiation monitor room

<1K-2K 0.4 Grade level Compactor room

<lK-6K 1.0-5.0 Solidification room

<1K-3K Hot machine shop

<1K-2K 0.06-0.1 f t level Miscellaneous waste filter room 2K-500K 2.0-50.0 Deborating ion exchange room

<1K 0.01-0.4 West decay heat cooler room

<1K-10K 0.5-12.0

Table 6 (CONT'D)

Auxiliary Buildina Contamination levels and Exposure Rates Contamination Exposure 2

Location (Room and/or Component) Level, dpm/cm Rate, mR/hr "A" high-pressure injection pump room

<1K-20K Spent resin tank / crud tank room IK-100K 5.0-100.0 Underground tank farm 1K-60K 1.0-500.0 Miscellaneous waste tank room

<1K 0.5-7000.

Miscellaneous waste concentrates tank room

<1K-80K 10.0-1000.

Pump alley (-27-ft level)

<1K

<1.0-10.

East decay heat cooler room

<1K-300K 4.0-60.

Waste gas decay tank room

<1K 0.04-0.2 "B" high-pressure injection pump room

<lK-22K

<1.0-22.0 Hallways

<lK

<l.0-6.0 Miscellaneous waste concentrator 2K-60K 2.0-18.0 Primary / secondary ion exchanger valve gallery

<1K-12K 2.0-140.0 Makeup pump room 2K-14K 1.0-10.0 ft level East decay heat pump room Main level 5K-20K 6.0-100.0 Mezzanine

<1K-2K 2.0-40.0 West decay heat pump room Main level

<1K-40K

<l.0-12.0 Mezzanine

<1K-40K 0.2-5.0 3.1.3.3 Fuel Storaae Buildina The walkway around the SFP is on the +40-ft elevation of the fuel storage building.

The exposure rate in this area is 0.5 mR/hr.

Contamination levels 2

ranging up to 80,000 dpm/100 cm were found [Ref. 3].

i Exposure measurements ranging from 6 mR/hr to 200 mR/hr were found in the upender pit.

Underwater contact measurements of corrosion products ranged from 200 mR/hr to 5 R/hr (Ref. 3].

3.1.3.4 Turbine Buildina On the +40-ft elevation of the turbine building, exposure rates were found to be less than 0.1 mR/hr. The highest measured exposure rate found was 0.16 mR/hr.

Removable contamination levels were found in general areas to be less than 1000 dpm/100 cm2 [Ref. 3].

, General area exposure rates of 0.1 mR/hr, and a contact measurement of 12 mR/hr on one of the feedwater heaters were found on the +20-ft elevation of the 2

turbine.

Removable contamination was found to be less than 1000 dpm/100 cm

[Ref. 3].

The polisher sump, condensate pit sump, feedwater pumps, and hotwells are on the grade level of the turbine building. Surveys of the sumps found them to be at

' ackg. um.d levels.

Contamination levels were found to be less than v

1000 dpm/100 cm2 [Ref. 3].

3.1.3.5 Areas Outside of Ma.ior Structures The tank farm is located on the northwest side of the reactor building.

In the area of the borated water storage tank (BWST), exposure rates were found that t

range from 0.5 mR/hr to 10 mR/hr. Contact exposures as high as 200 mR/hr were measured on BWST valves and pipes.

Removable contamination levels (excluding the tritium evaporator day tank sample sink enclosure ranged up to 50,000 2

2 dpm/100 cm ) were, in general, less than 1000 dpm/100 cm.

One high radiation area in excess of 100 mR/hr and another area averaging-8 mR/hr were measured in the 10SB.

Ip [this building, contamination levels were, in general, less than 1000 dpm/100 cm Ref. 3].

3.1.3.6 Inaccessible Structures / Systems IRef. 31 A number of plant areas are inaccessible because of high radiation levels, or because of some structural barrier.

The estimated exposure rates for inaccessible areas are given in Table 7.

1 Table 7 Exposure Rates for Inaccessible Structures / Systems [Ref. 3]

Exposure last Range Location Access to Area mR/hr.

1 Reactor cavity 2/89 100-1500 RC drain tank 1987 5000-j0000 2

Demineralizer 1974 Spent resin tank 2/90 5

Radwaste crud tank 1974 20-30 Flash tank 1987 100-200 6

Letdown filters 1985 300-10p0 Backflush tank 1974 Reactor building 8

Emerg. sump 1988 2000 t

. Table 7 (CONT'D)

Exposure Rates for Inaccessible Structures / Systems [Ref. 3]

Exposure Last Range Location Access to Area mR/hr.

Reactor building Normal sumps 200-500' Decay heat pump room

& radioactive waste sumps 1987 3-20 10 surveys from 1986 indicate a levet of 1800 mR/hr on the vessel exterior at the ft level.

2 spent fuel, primary, and secondary system cubletes. The licensee compared the radiological status with isotopic concentrations of the resins.

Radioactivity of the resin concentration in the range of 3 to 4 gol/mt.

The resins were transferred into a high-integrity container in June 1990; the licensee anticipates transferring approximately 100 f t* of resin per year from the SFP and redweste waste system.

The licensee measured 1 R/hr to 10 R/hr through an opening in the filter shield wall.

Last survey when shielding was removed from filters for modifications.

Estimated that exposure rates are in the several R/hr range.

sump surveyed with 4 in, of water in the stsrp; removable contamination in excess of 500,000 dpm/cm'.

Estimate with 0.25 in, of water in sistps "A" and "B."

Debris cleanout of the sump had exposure rates of up to 1 R/hr.

3.1.3.7 Plant Radiation and Contamination levels Summary The staff evaluated the radiation hazards at RSNGS, based on the radionuclide inventory in Section 3.1.2, and the existing radiological contamination levels in the plant, as described in Section 3.1.3.

General area radiation exposure rates in the reactor building range from 0.2 to 500 mR/hr.

Hot-spot exposure i

rates associated with the RCS were found to range between 15 mr/hr to 5 R/hr measured in the general area.

The highest general area exposure rates in the auxiliary building are found on the ft and ft elevations. These elevations contain most of the contaminated rooms and cubicles, and the general area exposure rates range from

)

0.01 mR/hr to 7000 mR/hr.

Contact exposures as high as 50 R/hr were measured.

In the fuel storage building on the walkway around the SFP, general area exposure rates of 0.5 mR/hr exist, and exposure measurements in the range of 6 mR/hr to 200 mR/hr were found in the upender pit. Underwater contact exposure rates of corrosion products on the bottom of the SFP were measured at levels up to 5 R/hr.

)

i Ne gueral area of the turbine building had exposure rates greater than 0.2 mR/hr.

The maximum contact exposure rate found was 12 mR/hr on one of the feedwater heaters.

Because of high radiation levels, a number of structures and systems are inaccessible. The exposure rates for these areas are expected to be in the range of several R/hr or greater.

The available exposure rate estimates (based en %ccglete data) are summarized in Table 7.

The licensee used the " Quick-Look" and Phase I Characterization Program

[Ref. 20], and PNL-5146 [Ref. 23] as documents forming the' bases for estimated exposure rates at RSNGS. The " Quick-Look" and Phase I Characterization satisfy 10 CFR 20.201 requirements to survey the facility for radiological hazards. The staff considers that the information from the " Quick-Look" and Phase 1 Characterization summarized in the licensee PDP [Ref. 3] is reasonable and acceptable.

3.2 Radioloaical Protection Radiological controls at RSNGS are under the direct authority of the radiation protection superintendent, who reports to the nuclear plant closure manager.

The radiation protection superintendent maintains oversight of radiation control measures at RSNGS, and ensures compliance with radiation protection standards.

The radiation protection superintendent is also responsible for the day-to-day management of the following:

(a) contamination control programs, radioactive materials handling, and protection of workers against ionizing radiation; (b) training on radiation safety; (c) evaluation of the facility radiological status; (d) making recommendations for controlling or eliminating radiation hazards and controlling the radiation exposure of individuals; (e) the procurement, calibration, maintenance, and distribution of portable radiation monitoring instruments; and (f) directing traditional control operations, determining acceptable personnel exposure, maintaining accurate dose records, and enforcing observance of radiation protection standards [Ref. 3].

The licensee commitment to ALARA principles is implemented with the RPP.

During decommissioning, the licensee has committed to continue to maintain an ALARA program, in accordance with 10 CFR Part 20, Regulatory Guide 8.8, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable," [Ref. 26] and Regulatory Guide 8.10, " Operating Philosophy for Maintaining Occupational Radiation Exposure As Low As Is Reasonably Achievable" [Ref. 27].

The licensee management commitment to ALARA is contained in the USAR [Ref.1], and is implemented by the station administrative procedures, in technical specifications [Ref.14] and radiation protection / chemistry department implementing procedures.

The licensee ALARA methods, techniques, and practices implemented during decommissioning will-be the same as those used during plant operations [Ref. 3].

' Engineering controls will be used to reduce radiation exposure.

HEPA filter ventilation units are available for use in controlling potential or actual areas of airborne radioactivity. During decommissioning, the following are some of the anticipated changes:

(1) ALARA reviews and goals will remain the same, except that the organizations may change, prompting a change in the responsible individual (s), and (2) as work involving exposure decreases, the ALARA coordinator will receive support, as needed, from department managers [Ref. 3].

The plant closure manager will be responsible for decommissioning activities at RSNGS, and is then the manager responsible for oversight of the radiation protection control measures at the plant. The radiation protection department establishes and ensures compliance with radiation standards that include the following:

(a) providing protection for all persons against ionizing radiation, supervising the contamination control program, radioactive materials handling, and ensuring plant compliance with the appropriate Federal and State regulations; (b) ensuring that individuals are trained in radiation safety; (c) evaluating and reviewing the radiological status of the station; (d) making recommendations for the control or elimination of radiation hazards and controlling the radiation exposure of individuals, to maintain exposure ALARA; (e) procuring, calibrating, maintaining, and distributing portable radiation monitoring int,truments; and (f) directing radiation control operations, determining acceptable personnel exposure, maintaining accurate dose records, and enforcing observance of radiation protection standards [Ref. 3].

The licensee administrative controls include procedures for radiation work permits (RWPs), waste shipments and disposal, access control, and the control of radioactive releases.

Before releasing radioactive material to the environment, the licensee reviews the proposed releases, to ensure that they are in conform-ance with 10 CFR Part 20 and 10 CFR Part 50, for gaseous and liquid releases.

Proposed shipments of radioactive materials are reviewed to provide assurance that the shipments meet the requirements of 10 CFR Parts 71 and 61, and 49 CFR

[Ref. 3].

I Personnel entering an RCA wear thermoluminescent dosimeters (TLDs) and direct-reading dosimeters.

Personnel who require more extensive monitoring are given extremity TLDs and dosimeters, alarming dosimeters, or multiple whole-body TLDs.

lhe licensee will maintain exposures within the limits established in Part 20

[Ref. 3].

As a part of its RPP, the licensee has committed to continue to perform onsite surveys, to ensure that licensed materials are surveyed in accordance with technical specifications. Surveys will include radiation and contamination surveys for RWPs, updating posted areas, and free release of plant equipment.

The frequency of surveys will be in accordance with approved procedures

[Ref. 3].

l

.I

i

The licensee procedures for free release of materials are based on the guidance in IE Circular 81-07, " Control of Radioactive Contaminated Materials" [Ref. 28],

and NRC Information Notice 89-92, " Survey of Waste Before Disposal from Nuclear Reactor Facilities" [Ref. 29]. Materials removed from RCAs are monitored to2 verify that they have less than 1000 disintegrations per minute (dpm)/100 cm loose beta-gamma contamination.

The licensee monitors materials for alpha-contami-nation only if the materials are being removed from areas known or suspected to ho alpha-contaminated [Ref. 3].

The staff has reviewed the licensee RPP and the licensee commitment to ALARA principles, to understand the licensee radiological protection of the workers and the public during decommissioning.

The staff has also reviewed the licensee RPP, as documented in the PDP [Ref. 3], and responses to questions (Refs. 6, 7, and 8].

The staff found the commitment by the licensee to keep in place the ALARA program used during plant operation acceptable.

In a similar fashion, the licensee RPP, based on the staff previous approval of the plan, as documented in

[Ref. 2], was found acceptable.

4.0 RADI0 ACTIVE WASTE MANAGEMENT The licensee radioactive waste management and waste disposal activities during the preparation for Custodial-SAFSOR, preparation for Hardened-SAFSTOR, and Hardened-SAFSTOR periods were evaluated.

Large quantities of liquid and solid radwaste will be processed during the preparation period for Custodial-SAFSTOR.

The licensee intends to drain systems and remove ion-resins and filters, during the layup and preparation for Custodial-SAFSTOR periods.

The licensee anticipates that radioactive solid waste production will be reduced during the Custodial-SAFSTOR period, and be further reduced during Hardened-SAFSTOR.

Radioactive waste in all forms (gaseous, liquid, and solids), generated during

~

the preparation for Custodial-SAFSTOR, will be processed using existing plant radioactive waste systems [Ref. 3].

Detailed deferred-DECON waste processing issues are beyond the scope of this SAFSTOR decommissioning plan and will be addressed in the licensee updated Deferred-DECON decommissioning plan.

The licensee plans to store waste generated during the SAFSTOR period onsite until it can be shipped to a commercial low-level radioactive waste disposal site or a licensed decontamination and volume-reduction facility.

4.1 Liouid Radioactive Waste The RSNGS liquid radioactive waste system is described in Section 11.2 of the SER prepared by the staff [Ref. 2], and is summarized here.

The liquid radioactive waste system consists of the coolant radwaste system and the miscellaneous radwaste system. The coolant radwaste system function was to process boric acid used in the RCS.

The coolant radwaste system consists of storage tanks, pumps, filter demineralizers, and the boric acid evaporators.

The licensee intends to abandon the use of the coolant radwaste system during

)

the preparation for Custodial-SAFSTOR.

-i

, The miscellaneous radwaste system is used to collect radioactive waste water and process it.

Processed water (which contains tritium) is primarily used as makeup water for SFP and miscellaneous controlled area uses. A portion of the water is discharged via the regenerate holdup tanks and retention basins. Major inputs to the miscellaneous radwaste system are (a) auxiliary building sumps, (b) radiochemistry laboratory drains, (c) floor drains, and (d) polishing demineralizer sumps, when contaminated liquids are detected in the secondary ystems.

The major equipment in the miscellaneous radwaste system comprises various tanks and sumps for collecting and storing of liquid waste, waste evaporator and condensate tanks, and the blender / dryer that is used to solidify the concentrated bottoms from the evaporator [Ref. 3]. A complete description of the RSNGS liquid radioactive waste system is given in the RSNGS USAR [Ref. 1].

Radioactive liquids will be processed using a combination of filtration, demineralization, evaporation, solidification, and offsite incineration, during the Custodial-SAFSTOR and Hardened-SAFSTOR periods [Ref. 3].

For the entire i

preparation for Hardened-SAFST0R, the miscellaneous liquid radwaste system will be used, as needed, depending on the composition and volume of water to be processed.

During the Hardened-SAFSTOR period, the miscellaneous ~ liquid radwaste system will be modified to handle expected source volumes of water and anticipated activities levels.

The expected sources of waste water during Hardened-SAFSTOR will be from decontamination and from rainwater that leaks into the building.

The liquid radioactive waste system will be modified according to established I

procedures for collecting, transferring, and storing water from building sumps during the Hardened-SAFSTOR period. The water collected will '.e evaluated, 3

using the ODCM to determine the appropriate way to handle the rain water that leaks in [Ref. 3].

l The staff reviewed the licensee proposed operation of the liquid radioactive waste-treatment system, the radionuclide concentrations expected, and the release procedures that will be used. The licensee liquid radioactive waste treatment method is acceptable.

The staff based its acceptance on the licensee description of proposed treatment methods in its PDP [Ref. 3], and on the staff evaluation of the liquid radioactive waste treatment system documented in the staff SER [Ref.2].

4.2 Gaseous Radioactive Waste During the SAFSTOR period, it will be necessary to process airborne particulates. The plant uses HEPA filters in the ventilation exhaust system (s) to process airborne particulates.

. 1

. i Tritium released from the spent fuel in the SFP is the current source of gases released from RSNGS, and this will continue to be the source of future releases.

No other gases or particulates have been detected in the plant gaseous waste stream since 1989.

The licensee estimated that for the preparation for the Custodial-SAFSTOR period (April 1992 - June 1993), a total of 25 curies of tritium would be released, and approximately 40 curies would be released during the Custodial-SAFSTOR period (June 1992 - June 1995) [Ref. 6].

The licensee has committed to continue to operate and maintain the auxiliary building exhaust system for the entire SAFSTOR period. One exhaust fan will be operated continuously, and a second will be on standby. The exhaust from the auxiliary building will be monitored by the auxiliary building stack radiation monitor, and will be backed up by a sampling capability should the auxiliary building stack-monitor fail [Ref. 6].

The spent fuel building will continue to be ventilated by the auxiliary building exhaust system.

The exhaust fan will maintain-a slight negative pressure on the spent fuel building, to prevent the release of radioactive materials to the environment through unmonitored release points.

The licensee will modify the reactor building exhaust ventilation system. The modification will entail tieing the suction side of the reactor building ventilation system to the auxiliary building exhaust fans (Ref. 7].

The auxiliary building exhaust system is equipped with HEPA filters, and will be operated, tested, and maintained in accordance with a periodic maintenance and test program, and monitored in accordance with the ODCM.

The instrumentation setpoints and calculation methods specified in the ODCM will be used to ensure that the limits of 10 CFR Part 20 are not exceeded.

In addition, the staff has evaluated the licensee PDP [Ref. 3] and responses to questions [Ref. 6] and has determined that the gaseous waste treatment system continues to meet the applicable requirements of 10 CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64.

The staff, therefore, finds the gaseous waste treatment system acceptable.

4.3 Solid Radioactive Waste The licensee estimated that the volume of solid waste for the entire decommissioning process (i.e., preparatiop (of Custodial-SAFSTOR through 3

Deferred-DECON) is expected to be 7391 yd 199,557 ft ) [Ref. 3]. Tables 8 and 9 summarize the waste volumes for the SAFSTOR phase and the Deferred-DECON phases.

3 3

3 The licensee estimated that 198 yd (5346 ft ) of Class A waste, and 6 yd 3

(162 f t ) of Class B waste would be generated while RSNGS is in SAFSTOR (i.e.,

total volumes generated during preparation for Custodial-SAFSTOR through Hardened-SAFSTOR dormancy).

Potential sources of greater than Class "C" waste are the reactor vessel internal components and non-fuel irradiated components stored in the SFP [Ref. 6].

. The solid waste treatment methods proposed for use at RSNGS are acceptable. The staff based its acceptance on the documented staff evaluations in Section 11.5 of the staff SER [Ref. 2], and on its evaluation of. solid waste generation and processing at RSNGS, during the SAFSTOR phase of decommissioning.

Solid waste processing during the Deferred-DECON phase will be evaluated on the basis of information provided in an updated decommissioning plan for the Deferred-DECON period.

Table 8 Total Estimated Volume of Solid Radioactive Waste Generated During SAFSTOR Decommissioning Phase Waste Volume, yd*

Preparation for Custodial-SAFSTOR Soil remediation 13 Process liquid 6

Contaminated solid 73 Custodial-SAFSTOR' Contaminated solid 15 Preparationforyardened-SAFSTOR Contaminated solid 73 Hardened-SAFSTOR Contaminated solid 28 Preparation for DECON Contaminated solid 109 3

Total 317 The Custodial-SAFSTOR period is estimated to be 3 years.

The Hardened-SAFSTOR period is estimated to be 9.3 years.

In the event an IFSFI is not approved, the total waste voltsne Will not be greater than the amount that 4s shown here.

Table 9 Total Estimated Volume of Solid Radioactive Waste Generated During Deferred-DECON [Ref.6)

Waste Volume, yd' i

Spent fuel rack removal 470-Reactor coolant piping 56 Pressurizer relief tank 13 Reactor coolant pumps & motors 412 Pressurizer 159 Steam generator 1036 CRDMs/lC1s'/ service structures removal 72 Reactor vessel internals 204 Reactor vessel 277 Coolant radwaste system 302 Decay heat removal 541

4 e Table 9 (CONT'D)

Total Estimated Volume of Solid Radioactive Waste Generated During Deferred-DECON [Ref.6]

Waste Volume, yd*

HVAC' auxiliary building 403 2

HVAC reactor building 530 Miscellaneous radwaste liquid 221 Radiation monitoring 36 Reactor building spray 110 Reactor coolant & chemical addition 38 Spent fuel cooling 119 Waste gas 39 Electrical (contaminated) 850 Secondary building 7

Reactor building 432 Auxiliary building 103 fuel storage building 80 Process liquid waste 122 Contaminated solid waste 206 Makeup & Purification 236 Total 7074 Control rod drive mechanisms /incore instrtanentation.

Heating, ventilation, and air conditioning.

4.4 Process and Effluent Radiolooical Monitorino Systems The ODCM will continue to be used, during decommissioning, to control releases of radioactive materials through gaseous / airborne pathways-.

Instrumentation, setpoints, and calculation methodologies specified in the ODCM are intended to ensure that the concentration limits of Appendix B, Table II, Column 2, to Part 20 (25 FR 10914, November 17,1960) are not exceeded, and the operability and use of instrumentation are consistent with the requirements of " General Design Criteria for Nuclear Plants" (GDC 60, 63, and 64 of Appendix A to 10 CFR Part 50). The annual dose commitments to an individual in an unrestricted area from radioactive materials in liquid and gaseous effluents meet the design objectives of Appendix I to 10 CFR Part 50. The staff finds acceptable the process and effluent monitoring methods that the licensee will use, during decommissioning at RSNGS.

5

. 4.5 Spent fuel Disposition Spent fuel disposal is not considered part of the decommissioning process (53 FR 24028, 24019, June 27, 1988) [Ref. 30].

Currently, there are 493 spent fuel assemblies, which contain approximately 140,800,000 curies, stored in the SFP at RSNGS. Spent fuel will remain stored in the SFP during the Custodial-SAFSTOR period.

If there is slippage in the licensee-proposed schedule for transferring the spent fuel to an onsite ISFSI, the Custodial-SAFSTOR period

+

will t>e extended until all Part 72 ISFSI issues are resolved or the DECON period is entered.

Slippage of the licensee schedule has no significant impact on the selected decommissioning alternative, because continued storage of spent fuel in the-SFP is acceptable for up to 30 years in accordance with 10 CFR 51.23(a),

i 4.6 Waste Handlina and Packaaina The licensee committed to take the following appropriate steps for preparing hazardous materials for shipment:

Identify the waste materials.

Determine the waste class.

Determine the proper shipping name.

Ensure proper packaging of the waste.

i Prepare the proper shipping papers.

Include the required certifications.

Mark and label the packages.

Load, block, and brace the loaded waste.

Apply the proper placards.

All waste will be shipped in accordance with the applicable requirements.

The staff has reviewed the licensee plans.for waste handling and packaging and concludes that they are consistent with the applicable provisions of 10 CFR -

Parts 20, 61, 71; Department of Transportation requirements; and the staff

" Technical Position on Waste Form" [Ref. 31]; therefore, the licensee plans are acceptable.

4.7 Waste Transportation The licensee addressed radioactive waste transportation issues in its PDP

[Ref. 3].

The staff has reviewed the licensee plans for waste transportation and concludes that they are consistent with the applicable provisions of 10 CFR Parts 20, 61, 71; Department of Transportation requirements; and the staff

" Technical Position on Waste Form" [Ref. 31]; therefore, the licensee plans are acceptable.

5.0 FINAL RADIATION SURVEY PLAN The final radiation survey will take place after the Deferred-DECON. The licensee will update its decommissioning plan to address the Deferred-DECON decommissioning, in accordance with NRC guidelines to address residual contamination levels before release of materials and the facility for unre-stricted use.

i 6.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS IN PLACE DURING DECOMMISSIONING The staff reviewed the licensee PDTS that would be in place during decommis-sioning. The PDTS incorporated in Amendment No. 119 of the facility license

[Ref 14] and Possession Only License (POL) (Amendment No. 117 effective April 28, 1992), will be in place during decommissioning.

7.0 OUALITY ASSURANCE j

The licensee Quality Assurance Program (QAP) is defined in the Rancho Seco SAFSTOR Quality Manual (RSSQM). The RSSQM addresses provisions during the Custodial-SAFSTOR period. The purpose of the RSSQM is to ensure that decommissioning and other specified activities are performed, during Custodial-l SAFSTOR, in accordance with the plant license, applicable codes and standards, and regulatory requirements [Ref. 3].

The QAP for the Custodial-SAFSTOR phase of SAFSTOR will include the following:

Definition of the quality and QA objectives and requirements Assignment of organizational responsibility and authority for activities affecting quality Identification of systems, regulatory-based programs, and decommissioning activities to which the QAP applies Requirements for implementation procedures and instructions that are appropriate for the activities to be performed Provisions for independent audits and reviews, verification, quality attainment, and QA effectiveness Means for prompt identification and correction of deficiencies The RSSQM addresses the criteria of 10 CFR Part 50, Appendix B, and describes how the controls pertinent to each are to be implemented.

7.1 OA Oraanization QA personnel assigned to RSNGS will have defined responsibilities, organizational freedom, and independence to identify problems that affect quality; to initiate, recommend, or provide solutions; and to verify implementation of solutions. The appropriate personnel in the QA organization have the authority to initiate actions necessary to correct problems and to bring any unsatisfactory condition into conformance.

l i

The nuclear plant CM is responsible for the safe and reliable decommissioning of RSNGS, and reports directly to the AGM. The AGM is responsible for development and execution of the RSNGS QA Plan.

For day-to-day operation, the QA organization reports to the CM.

QA will be independent of the pressures of other lines of authority. The AGM will maintain involvement in the QA matters, and will assess the scope, status, implementation, and effectiveness of the QAP through the NSRCA [Ref. 3].

7.2 OA Plan t

The QA Plan for RSNGS is described in RSSQM. The program for the Custodial-SAFSTOR is described.

The Custodial-SAFSTOR QAP will ensure the following:

Plant configuration is in accordance with the " Rancho Seco Defueled Safety Analysis Report" (RSDSAR).

Compliance exists with applicable license conditions and defueled technical specifications.

Facility programs, structures, systems, and components satisfy applicable regulatory requirements.

The RSNGS SAFSTOR QAP applies to structures, systems, and components that prevent or mitigate the consequences of accidents, analyzed in the RSDSAR, that could have an impact on the health and safety of the public.

Applicable portions of the SAFSTOR QAP will apply to programs for radioactive effluent control, radiological environmental monitoring, radiation protection, radiological waste handling and shipping (including process control, security, fire protection, training, and emergency preparedness) [Ref. 3]. The staff -

found the information provided by the licensee in its PDP [Ref. 3] acceptable, in accordance with 10 CFR 50.82(b)(5).

The staff evaluated the licensee QA provisions during decommissioning, concluding that the licensee program is acceptable, because it meets the applicable sections of NUREG-0800 [Ref. 18), Section 17.2, " Quality Assurance during the Operations Phase."

8.0 POSTULATED ACCIDENTS I

The primary impact of postulated decommissioning accidents would be the release of radioactive materials into the environment and the resulting exposure of the public to radiation.

other tun those evaluated]icensee has not identified any additional accidents, The in NUREG/CR-0130 [Ref. 12], that would have an impact on the public. The radionuclide inventory at RSNGS is less than the 3 See Table 11.1-2, in NUREG/CR-0130 [Ref.12] for a summary of accidents and releases.

Table 11.2-3 contains a summary of the higher-consequence accidents postulated for immediate dismantlement and preparation for SAFSTOR, and associated calculated doses.

r i

- = -

radionuclide inventory at the reference PWR evaluated in NUREG/CR-0130

[Ref. 12]. This supports the licensee contention that no significant exposure to members of the public will result from onsite accidents [Ref. 3].

The licensee evaluated the following three additional accidents for the SAFSTOR period:

(a) fuel handling accident, (b) complete loss of offsite power, and (c) the drop of a spent fuel cask [Ref. 3].

Because of the mechanical and electrical interlocks and the administrative controls associated with the turbine building gantry crane, which prevent movement of a cask over spent fuel assemblies in the SFP, dropping a cask onto spent fuel is not considered a credible accident.

The fuel-handling accident and the complete loss of offsite power are two accidents that are applicable during SAFSTOR.

l The environmental effects of the fuel-handling and the loss-of-offsite-power accidents, in terms of the radiation exposures to potential receptors, are listed in Table 14.3-1 of the RSNGS USAR [Ref.1]. The short-lived radionuclides in the spent fuel have decayed substantially since the plant was shut down on June 7, 1989.

Therefore, the dose impact is significantly less than the dose for an operating plant.

The total calculated thyro'id dose from a credible accident in the permanently defueled mode (PDM) is less than 1 mrem

[Ref. 3].

The raiionuclide of concern for a fuel-handling accident is Kr-85.

On the basis of the licensee analysis of this accident, in the current condition, a 2-hour integrated total body dose attributed to the maximumally exposed individual is 0.013 rem.

This dose is a small fraction of the 10 CFR Part 100 accident dose limit of 25 rem. The 13 mrem whole-body dose is approximately 1.3 percent of the 1000 mrem protective action guideline (PAG) recommended by the Environmental Protection Agency (EPA), "A Manual of Protective Action for Nuclear Incidents,"

EPA 520/1-75-001 [Ref. 32].

The health and safety impacts of potential decommissioning accidents are considered to be acceptably low.

During a complete loss of offsite power, control of the spent fuel decay heat load, and thus the protection of the spent fuel integrity, is the primary consideration.

The controls required to protect the spent fuel are based on anticipated decay heat generated by the spent fuel stored in the SFP. A complete loss of offsite power would result in _ loss of the spent fuel cooling system (SFC) and of the spent fuel building ventilation system.

SMUD analyzed the effect of a complete loss of offsite power on the SFP [Ref. 38]. This analysis assumed the most limiting initial conditions allowed by the RSNGS technical specifications; i.e., the SFP water level as low as 23 feet 3 inches and the SFP bulk water temperature as high as 140 degrees Fahrenheit.

The normal operating conditions of the SFP are the water level above 37 feet and the bulk water temperature below 90 degrees Fahreheit. The results of the analysis for an initial temperature of 140 degrees and the initial water level of 23 feet and 3 inches indicated that it would take over 15 days, after bulk boiling begins, for boiling to reduce the level of water in the SFP to the top of the spent fuel assemblies. This analysis did not take credit for any convective cooling by the ventilation system. This period of 15 days did not include the time required to raise the SFP bulk water temperature from its initial a

1

  • temperature of 140 degrees Fahrenheit to 212 degrees Fahrenheit, the boiling point of water at standard conditions.

Boiling of the water will not damage the fuel assemblies, because these fuel assemblies are designed to' function at temperatures much higher than 212 degrees Fahrenheit.

The staff, using very conservative assumptions, calculated that it would take at least 3 days to increase the SFP bulk temperature from 140 degrees Fahrenheit to 212 degrees Fahrenheit, with the initial SFP level at 23 feet and 3 inches.

The assumptions the staff made to calculate SFP heat-up time to reach the bulk boiling temperature were the following:

(1) The energy addition rate to the SFP water was assumed to be 1.68E6 BTV/HR.

This enprgy rate was based on the decay energy in the fuel as of November 1, 1991. The current decay energy of the fuel is actually less than the value used because of the additional 1% year-period the fuel has been stored in the SFP. The use of the November 1, 1991, decay rate artificially increased the amount of energy assumed to be added to the SFP water.

This assumption was conservative because the net effect was to increase the rate the SFP temperature was raised.

(2) The boundary of the water volume in the SFP was assumed to be perfectly insulated.

That is, no energy was allowed to be transferred i

from the SFP water volume to (or through) the walls, floor, or surface of the SFP.

The energy that would have escaped from the SFP water was assumed artificially to be absorbed by the SFP water, increasing the SFP water temperature. This assumption was conservative because the net effect was to increase the rate the SFP temperature was raised.

(3) No credit was given for the energy absorption capability of the metal located in the SFP (fuel storage racks).

This metal would act as a heat i

sink during a SFP water heat-up evolution (from 140 degrees Fahrenheit to 212 degrees Fahrenheit).

The energy that would be absorbed by the metal

~

was artificially assumed to be absorbed by the SFP water and increased the SFP water temperature. This assumption was conservative because the net i

effect was to increase the rate the SFP temperature was raised.

(4) The volume of the fuel and fuel racks were assumed to displace 20 percent of the water in the SFP. This displaced volume was not included in the calculation, reducing the mass of SFP water available to absorb the energy added to the SFP water. This assumption increased the amount of energy absorbed per unit mass of SFP water, which increased the rate the SFP water temperature was raised.

i i

' Sacramento Municipal District Utility letter to NRC, dated November 19, 1991, Attachment II at page 54.

1 I

(5) The loss of water mass from the SFP due to evaporation during the heat-up evolution was not included in this calculation.

It was the technical judgment "of the staff that the added complexity to model this evaporation phenomenon (mass and associated energy loss) was not necessary because the net effect on the heat-up time would be minimal due to the counteracting effects of mass loss and associated energy loss. With each unit of SFP water mass lost due to evaporation, an associated amount of energy will also be removed from the remaining -SFP water volume.

The net effect of mass and energy removal would be to reduce the rate at which the SFP water temperature would rise.

Additionally, the licensee calculated that approximately 1 foot of SFP water level would evaporate every 70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> if the SFP water temperature was at a steady-state temperature of 180 degrees Fahrenheit. This steady-state temperature is approximately the mean temperature for the heat-up evolution considered.

This further supports the decision not to include the evaporation phenomenon because of the minimal effect on the time to raise the SFP water temperature from 140 degrees Fahrenheit to 212 degrees Fahrenheit.

SMUD also has procedures that address loss of offsite power to RSNGS.

SMUD has-equipment onsite (a diesel powered fire pump) which may be used to add water to the SFP, if necessary, even during periods when offsite power is unavailable.

The 18-day period provides ample time to take corrective action even with a complete loss of offsite power.

The staff has concluded that the complete loss of-offsite power to RSNGS during decommissioning will not significantly impact the health and safety of the public.

The staff bases its conclusion on the considerable length of time available for SMUD to implement its loss of offsite power procedures to either restore offsite power or take corrective measures such as adding water to the-SFP with existing equipment which does not need offsite power to function.

Accidents that could occur during the Deferred-DECON are outside the scope of this SER, The staff has reviewed the licensee postulated accident scenarios and found that none of the accidents has potential coasequences (radiation doses) in excess of PAG levels recommended by the EPA in "A Manual of Protective Actions for Nuclear Incidents," 1991 [Ref. 32].

9.0 FINANCIAL ASSURANCE 1

9.1 Cost Estimates for Decommissionina The review of the cost estimates for decommissioning RSNGS was based on independent estimates and a comparison of several activities to be conducted at i

this facility to similar activities conducted at other facilities.

The staff review included an evaluation of the licensee cost assumptions used for major decommissioning activities and tasks, including:

a review of the deferred dismantlement and decontamination costs, waste disposal costs (based on estimated volumes of waste generated), transportation costs, equipment costs,

)

. a-d 1:bar rates.

The staff evaluation of the RSNGS decommissioning cost was based on comparison methods used to evaluate decommissioning cost estimates for Pathfinder, Fort Saint Vrain (FSV), and Shoreham.

In addition, the "1992 Means Building Construction Cost Data" [Ref. 33), the " Dodge Manual for Building Construction Cost Data," 1984 [Ref. 34], and NUREG/CR-0130 [Ref.-12], were used by the staff in its analysis. All cost information was escalated to 1991 dollars, using an annual inflation rate of 5 percent. The estimated cost of

b=t !280,000,000 represents an acceptably conservative estimate of decommissioning the RSNGS facility.

This $280,000,000 figure includes spent fuel storage and " green field" restoration costs which are not considered to be decommissioning costs under NRC regulations.

Many activities and tasks that will be necessary to place RSNGS into a SAFSTOR condition are similar to activities and tasks conducted at other reactor plants that have been decommissioned. The staff reviewed several areas to ensure that the estimated cost to eventually decontaminate the RSNGS is reasonable. For example, the cost of removal of contaminated pumps (1,000-10,000 lbs.) was compared to a similar activity that was conducted at the Pathfinder facility.

j The removal of similar pumps (1,000-10,000 lbs.), at Pathfinder, cost approxi-mately $1900. The cost of removing similar pumps at FSV is estimated to be

$3065.

The estimated cost to remove similar-size contaminated pumps at RSNGS is estimated to be $4364.

Even after adjustments for regional differences and inflation, the RSNGS costs were greater than the estimated co;ts at Pathfinder and FSV.

To date, the actual costs for decommissioning at Pathfinder have been consistent with the initial estimate, and, therefore, represent an actual basis for comparison.

The staff compared the labor rates summarized in the RSNGS Decommissioning Cost Estimate, Section 2, " Unit Cost Factors," for RSNGS to the estimated labor rates for Pathfinder, FSV, and Shoreham, using the city cost indexes in "1992 Means-Building Construction Cost Data" [Ref. 33] and an escalation of 5 percent, and found them reasonable. Also the staff compared the work difficulty factors used in the R$NGS cost estimate, for adjusting activities to accommodate for access adjustment, radiation, respiratory protection, and work breaks, to the work difficulty factors used for Shoreham and Pathfinder cost estimates, and concluded that they are reasonable.

The staff compared RSNGS estimated equipment rental costs to the cost for i

equipment rentals listed in "1992 Means Building Construction Cost Data"

[Ref. 33], and adjusted the costs, using the city. cost index.

It examined the rental costs of many different types and sizes of equipment, ranging from small air compressors to 50- to 70-ton cranes.

The estimated equipment costs were based on the cost estimates for equipment found in "1992 Means Building Construction Cost Data" [Ref. 33], and are' reasonable estimates for the cost to rent equipment.

3 The estimated costs of removing piping systems at RSNGS were compared to those at Pathfinder and FSV.

For example, the estimated cost of removing piping from 2 to 8 inches in diameter, at RSNGS, is $67.25 per foot, compared to the estimated cost of removing piping up to 5 inches in diameter, at FSV, for $74.83 per foot.

The estimated cost of removing 2-to 8-inch-diameter piping at Pathfinder was $30.54 per foot. The estimated cost for removing the FSV cooling

. water system piping, which consists of more than 47,000 feet of piping, ranging in diameter from 0.5 to 20 inches, averages $108.79 per foot.

The estimates cost to remove piping greater than 8 inches in diameter at Pathfinder was $60.36 per foot.

In comp:rison, the estimated cost to remove contaminated piping greater than 8 inches in diameter for RSNGS is $132.98.

After adjustments for inflation and regional differences, the RSNGS estimates were considered reasonable.

In addition, the staff compared the estimated cost of removing contaminated concrete, from the RSNGS, to the cost of removing contaminated concrete at FSV, and to the actual cost of removing contaminated reinforced concrete at Pathfinder.

Although the methods of removing the concrete are different, the cost of removal should be similar.

RSNGS estimated the cost of removing contaminated concrete at $1333 per cubic yard, FSV estimated the cost for removing contaminated concrete at approximately $1120 per cubic yard, while the cost of removing contaminated concrete at Pathfinder was estimated at $650 per cubic yard.

Using the Means City Cost Indexes. and adjusting for regional differences, the adjusted RSNGS cost is estimated at $1220 per cubic yard, compared to the adjusted FSV estimated cost of $1230 per cubic yard.

The actual adjusted cost of removing contaminated concrete for Pathfinder is $895 per cubic yard, after adjustment.

Therefore, the estimate to remove the contaminated concrete at RSNGS is reasonable. The staff also reviewed the cost for disposal of the 192,000 estimated cubic feet of radioactive materials and finds the estimated disposal cost of $35,144,000 low. However, RSNGS has included a 7

24 percent contingency for disposal. This results in an estimated disposal cost of 543,580,000, or a disposal cost of $220 per cubic foot.

Based on the current i

and projected disposal costs, this estimated disposal cost of approximately-

$220 per cubic foot, in 1991 dollars, is reasonable for Class A waste.

The staff concludes that this cost estimate for decommissioning the RSNGS facil-ity meets the requirement of 10 CFR 50.82(b)(4).

9.2 Method for Fundina Decommissionino i

In a letter of July 24, 1990, as supplemented on May 26 and July 19, 1991, the Sacramento Municipal Utility District (SMUD or the licensee) submitted its plan for the accumulation of funds for decommissioning and requested an exemption from 10 CFR 50.75(e)(1)(ii) concerning the requirement to have all decommissioning funds collected when operations are terminated. On November 13, 1991, the staff issued an exemption to SMUD from 10 CFR 50.75 (e)(1)(ii) and approved the Rancho Seco Funding Plan. However, on April 24, 1992, the NRC revoked this exemption and approval of the funding plan (57 FR 15117) because the staff became aware that the NRC had previously committed ~to seek public participation concerning regulatory relief regarding the Rancho Seco decommissioning.

Consequently, on May 14, 1992, the NRC published a Notice of Consideration of Issuance of Exemption and solicited public comments on the proposed exemption and approval of the funding plan (57 FR 20718). On June 15, 1992, the Environmental and Resources Conservation Organization (ECO), submitted comments in response to the staff's solicitation. On July 9,1992, the Final Rule on Decommissioning Funding for Prematurely Shut Down Power Reactors was-issued (57 FR 30383).

This rule, which became effective on August 10, 1992,

, amended the NRC regulations in 10 CFR 50.82, regarding the timing of collection of funds for decommissioning for prematurely shut down power reactors.

The amended regulation requires that the NRC evaluate decommissioning funding plans for these plants on a case-by-case basis, taking into account site-specific safety and financial situations.

In effect, this rule renders moot the need for an exemption from the earlier regulations. Therefore, the staff has undertaken an evaluation of the SMUD decommissioning funding plan under the amended regulation and will discontinue the consideration of an exemption for approval of this plan.

In addition, the evaluation below considers the public comments received from ECO on June 15, 1992, on the funding plan.

9.2.1 Financial Assurance Plan In its July 24, 1990, and March 26 and July 19, 1991 submittals, the licensee outlined its plan to ensure accumulation of decommissioning funding. The site-specific cost estimate submitted with the SMUD decommissioning plan of May 20, 1991, calculated the cost of decommissioning RSNGS to be about $280 million.

In this SER, the NRC staff addresses whether the SMUD plan for accumulating these funds is adequate to assure public health and safety.

The licensee plan to ensure the accumulation of the estimated $280 million l

required for decommissioning comprises the following:

i (1) initial contribution of approximately $90 million into an external sinking trust fund (2) annual contributions of approximately $8 million until the decommissioning plan is approved and $12 million annually thereafter 4

(3) annual contributions to continue until end of operating license period (2008) when funding will be complete In accordance with 10 CFR 50.82(a), the NRC evaluation of funding plans is to take into account site-specific safety and financial situations.

Additionally, the staff considered the comments received from ECO.

9.2.1.1 Safety Considerations On June 7, 1989, the RSNGS facility was shut down. The reactor was completely defueled on December 8,1989, and the fuel was stored in the spent fuel pool.

With Rancho Seco in a defueled condition, the probability of previously analyzed accidents occurring has been significantly reduced.

In addition, the radionuclide inventory at RSNGS is less than the radionuclide inventory at the reference PWR evaluated in NUREG/CR-0130 [Ref. 20]. As discussed earlier in Section 8, the accidents postulated for SAFSTOR at RSNGS are bounded by the accidents postulated in NUREG/CR-0130 [Ref. 20] which formed the basis of NUREG-0586 [Ref. 5].

Therefore, in light of the long-term defueled condition at R'Das and the negligible accident consequences of the spent fuel, the staff has concluded that there is no special or unusual safety concern at RSNGS that could negatively

.mpact the proposed accumulation of funds for decommissioning at Rancho Seco.

J

. 9.2.1.2 Financial Considerations The staff reviewed financial considerations at RSNGS to provide reasonable assurance that SMUD has the ability to collect the necessary decommissioning funds under its proposed plan. To this end, the staff evaluated (1) the licensee ability to make funding contributions and (2) the reasonableness of the accumulation schedule.

(1) Ability To Make Fund Contributions To assess the SMUD ability to make fund contributions, the staff considered such areas as (a) financial security, (b) history of. fund collection, (c) rate regulation, and (d) other decommissioning obligations.

(a)

Financial Security - SMUD has a Moody bond rating of "A."

This bond rating indicates that SMUD has a high degree of financial security and, therefore, provides assurance that SMUD decommissioning funding requirements can be met.

(b) History of Fund Collection - The licensee has made annual contributions to its decommissioning fund since January *1, 1980, even in periods of financial difficulty.

SMUD has already accumulated approximately $90 million of the estimated $280 million needed to decommission RSNGS.

SMUD is maintaining its commitment to set aside funds over a long period, thus providing additional confidence that funds for decommissioning RSNGS will continue to be accumulated.

(c) Rate Reaulation - As a municipal utility (unlike a privately owned utility) SMUD has legal authority to establish its own rates and charges. Therefore, it has the ability to recover costs associated with the RSNGS decommissioning from its ratepayers. The District resolution 91-6-9 provides that "the District will collect through rates the cost of SAFSTOR and decommissioning of $280 million...."

The NRC finds that the ability to establish rates is a key factor in ensuring decommissioning funding and, therefore, provides additional assurance that funds for decommissioning Rancho Seco can be accumulated.

(d) Obliaations To Decommission - Aside from _the decommissioning regulations imposed by the NRC, there are other laws applicable to SMUD that provide additional obligations regarding decommissioning RSNGS.

In a letter of March 18,-1991, to Walter Gaebler of SMUD, the law firm of Orrick, Herrington, & Sutcliffe stated an opinion that the Cortese-Knox Local Government Act of 1985, California Government Code, s 56000, e_t seo., (the Act) generally provides for the continuation of the decommissioning obligations to the NRC in a successor to SMUD in the event of a dissolution of SMUD.

The opinion states that the Act contains detailed information as to the provisions relating to a dissolution. These include such areas as:

(1) providing that the successor has the same powers and duties as the dissolved body; that

t

{ t (2) any monies and funds from the dissolved body and monies and funds.

received by the successor shall be used for payments due on bonds and other contracts or obligations; and (3) the rights of the successor are subject to the provisions of contracts or other obligations.

The opinion concludes that the Act would require any successor to SMUD to i

continue to fulfill SMUD duties and to apply SMUD revenues to the payment of the District obligations.

Additionally, the SMUD General Counsel, in an opinion of June 19, 1991, concluded that the California Nuclear Facilities Decommissioning Act, California Public Utilities Code, f 8325, places an obligation on SMUD to provide adequate funds for decommissioning.

This act has provisions for decommissioning items such as:

(1) providing for an estimate of decommissioning costs; (2) providing for the establishment and management of a separate external fund for the purpose of nuclear facility decommissioning; (3) providing for the periodic revision of j

the decommissioning cost estimate; and (4) developing regulations related to realistic cost estimating, periodic reviews, and cost controls.

These two State acts offer additional assurance that funds will be available to decommission the facility.

(2) Reasonableness of Accumulation Schedule i

The SMUD accumulation schedule for decommissioning funding is presented in l

Table 2 of the July 19, 1991, submittal, " Accumulation of Funds for Decommissioning." This schedule is based on the 5280 million site-specific cost estimate performed by TLG Engineering.

The funding plan calls for a 17-year active funding period until 2008 (license expiration).

The evaluation of the accumulation schedule by the staff considered (a) funding period, (b) initial contribution, (c) assumed investment and inflation rates, and (d) contributions and disbursals.

(a)

Fundina Period - SMUD intends to fund the Rancho Seco decommissioning-through yearly additions and earnings throughout the SAFSTOR phase of decommissioning. All the necessary funds for decontaminating RSNGS and returning it to unrestricted use (i.e., DECON or Dismantlement) will be accumulated by 2008, before DECON operations begin.

j The SAFSTOR phase of decommissioning represents a storage period in which radionuclides are allowed to decay over time, resulting in reduced generation of low-level radioactive waste.

The staff concludes that as long as all the necessary funds for DECON are accumulated before the DECON phase commences, funds may be permitted to accrue throughout the SAFSTOR phase of decommissioning.

= (b)

Initial Contribution - the SMUD accumulation schedule includes an j

initial contribution of approximately $90 million.

This is a substantial portion of the total funds required for decommissioning.

This initial contribution gives SMUD a solid financial footing for entering decommissioning in that it allows for high earnings at the outset and can be used to place the plant in a safe configuration in the event of unforeseen occurrences.

i (c) Assumed Investment and Inflation Rate - SMUD assumes an investment return of 8.3 percent annually. The licensee stated that its decommissioning trust fund investments will be primarily in U.S. Treasury and Government Agency securities and that securities will be purchased with "a buy and hold to maturity" philosophy.

The licensee decommissioning investments are conservative. The assumed investment rate of 8.3 percent for U.S. Treasury and Government Agency securities, however, is optimistic considering recent 30-year Treasury yields.

On the other hand, the licensee estimate for inflation of 5.1 percent assumed in the decommissioning cost estimate is correspondingly pessimistic.

Therefore, the real rate of return (nominal minus inflation) is reasonable and acceptable.

Future rates of return are sufficiently uncertain that the accumulation plan must be evaluated by the licensee at least biennially (every 2 years) and adjusted at least every 5 years.

This will allow adjustments to be made to compensate for variances in actual rates of return (or other changes to the decommissioning plan) that may negatively impact the total fund accumulation.

(d)

Contributions and Disbursals The licensee accumulation schedule includes annual contributions of approximately $12 million until 2008 (commencement of DECON).

Additionally, the schedule includes annual disbursements of $4 million to $40 million during the SAFSTOR phase of decommissioning and

$65 million to $140 million during the DECON phase of decommissioning.

The licensee contribution and disbursal schedule meets the annual funding requirements and total decommissioning cost estimate.

1 Additionally, the annual disbursements are reasonable and consistent with particularities of each phase of decommissioning. The staff notes, however, that the Table 2 earnings predictions rely on annual contributions and disbursements at the beginning of each year.

Therefore, as mentioned previously, the staff has determined that biennial reviews of the accumulation schedule by the licensee must be i

conducted (with adjustment at least every 5 years).

This will ensure that slight variances in disbursements and contributions can be corrected early before shortfalls compound to less manageable levels.

. 9.2.2 Eublic Comments

.As mentioned earlier, although an exemption is no longer required due to the rule change regarding decommissioning funding for prematurely shut down power reactors, the staff has considered public comments previously received.

On June 15, 1992, the Environmental and Resources Conservation Organization (ECO) submitted comments in response to the Notice of Consideration of Issuance l

ui Exempt wn w 5 MUD regarding decommissioning funding published in the Federal Reaister.

Some of the comments by ECO address the adequacy of the NRC Federal Reaister notice and the Environmental Assessment and Finding of No Significant i

Impact in support of an exemption to the NRC funding regulations. Although the RSNGS exemption request has become moot, the staff will address those few comments that raise a particular safety or financial concern regarding SMUD financial assurance plan for decommissioning.

(1) COMMENT: There is undue risk to the health and safety of the public that adequate funds will not be available for decommissioning when needed since Ranch Seco is no longer producing revenue and the SMUD financial stability is uncertain due to:

(a)

SMUD "BBB" bond rating (b)

SMUD investment in geothermal steam project (c)

SMUD involvement in lawsuits with industrial customers (d)

SMUD involvement in the California /0regon Transmission Project (C0TP) may cause them to acquire a large obligation (e) consumers groups petitioning against the SMUD $800 million bond filing STAFF RESPONSE:

(a)

ECO is incorrect in its characterization of the SMUD bond rating.. -I n fact, the SMUD bond rating has improved and is rated "A" by Moody's, "A " by Fitch, and "BBB+" by Standard and Poors.

Further, when the funding rule for prematurely decommissioned plants was issued the Commission stated that the NRC did not intend that this rule set a i

mandatory requirement that a minimum "A" rating must be met before the NRC would approve funding into a shut down reactor's safe storage period.

Rather, one reason the "A" rating criterion was proposed was to serve as a screening test of whether additional financial data was required to determine whether the licensee should be allowed to fund decommissioning into a storage period.

If the licensee met this criterion, the licensee would not have to prepare and submit additional documentation of its financial situation to be allowed to fund decommi.ssioning into a storage period.5 SMUD has provided information on its financial situation to justify being allowed to fund decommissioning into the SAFSTOR period.

This information was previously discussed in Section 9.2.1.2, Financial Considerations.

5 Federal Reaister (57 FR 30383), Analysis of and Response to Comments section of the final rule that revised 10 CFR Part 50.82(a).

. (b)

The SMUD investment in the geothermal steam project is a minor part of its operation and is already covered in rates.

(c)

The lawsuit with industrial customers was resolved and dismissed.

(d)

SMUD will incur an additional $4 million to $5 million per year from its involvement in COTP; however, increased capacity will more than offset this increase.

l (e)

SMUD withdrew its bond filing and replaced it with a request for a l

lower amount of authority which met no opposition.

(2)

COMMENT:

SMUD should be required to fund the Rancho Seco decommissioning by 1998 and before receiving a POL so that those who received the. benefit from Rancho Seco pay for the decommissioning rather than their children.

STAFF RESPONSE:

The POL was issued to SMUD on March 19, 1992, and became effective on April 28, 1992. Additionally, the staff has concluded that requiring a greatly reduced funding period is unnecessarily burdensome to the licensee.

However, to avoid situations where lack of funds could delay

[

and degrade the decommissioning process to the detriment of the public i

health and safety, the NRC maintains the policy that all funds be available prior to the start of the deferred DECON. Additionally, in the decommis-sioning rule (53 FR 24018, June 27,1988), the NRC stated its intent to defer to rate-making authorities on questions of intergenerational equity.

Consequently, the SMUD plan of funding through the SAFSTOR period is i

acceptable.

(3)

COMMENT: The decommissioning cost estimate is below that allowed by the -

NRC regulations.

l STAFF RESPONSE:

The cost estimate that ECO refers to was contained in the SMUD funding certification of July 24, 1990, and was based on the certification formula in 10 CFR 50.75(c). The current SMUD funding plan is based on a site-specific cost estimate provided with the proposed decommissioning plan of May 20, 1991, which exceeds the 10 CFR 50.75' certification formula.

(4) COMMENT: The interest earning rate is overly optimistic.

STAFF RESPONSE:

Although the SMUD assumed investment rate is optimistic, however, as. stated previously, since the SMUD. decommissioning-cost schedule assumes an overly high inflation rate, the real rate of return is reasonable. Additionally, the licensee has committed to biennial reviews and adjustments at least every 5 years so that any variances in investment return can be corrected.

il

. (5)

COMMENT:

The SMUD history of fund collection does not provide evidence that future collections can be made since SMUD generating capability has been reduced by one-half with the shutdown of RSNGS.

STAFF RESPONSE: The SMUD history of fund collection alone does not guarantee the SMUD capability of collecting funds in the future.

However, when combined with other indicators discussed previously, such as financial security, rate regulation, and other obligations to decommission RSNGS, there is reasonable assurance that SMUD will make future fund collections.

9.2.3 Decommissionina Fundina Summary Based on the foregoing evaluation of the SMUD funding plan, the staff concludes that none of ECO comments considered above offer any. evidence that the SMUD funding plan will be detrimental to the health and safety of the public.

Therefore, the staff concludes that the licensee plan for accumulating funds for-decommissioning is acceptable, provided that:

(1) biennial reviews of the accumulation schedule are conducted by the licensee, and (2) revisions are made to the accumulation schedule, not to exceed every 5 years, to reflect any variations in the decommissioning cost estimate, the decommissioning funding plan, or the decommissioning plan.

Further, the NRC finds that the licensee financial assurance plan offers adequate assurance that funds will be available to decommission RSNGS in a manner that protects the public health and safety.

10.0 STATUS OF DECOMMISSIONING PLAN LITIGATION On March 19, 1992, a Notice of Consideration of Issuance of an Order Authorizing Decommissioning a Facility and Opportunity for Hearing was published in the Federal Reaister (57 FR 9577), in accordance with the requirements of 10 CFR Part 50.82(e).

The Environmental and Resources Conservation Organization (ECO) in a letter of April 20, 1992, commented on the proposed action and requested a hearing. On May 8, 1992, the Secretary of the Commission forwarded the request for a hearing to the Atomic Safety and Licensing Board (ASLB) for further action.

The ASLB, in its Order LBP-92-23 dated August 20, 1992, (36 NRC 120, 1992) terminated the proceeding by denying standing to ECO.

On September 8, 1992, ECO appealed this ruling to the Commission pursuant to 10 CFR Part 2.714(a). The Commission exercising its authority granted discretionary r

standing to ECO and authorized ECO to submit amended contentions on (1) the i

adequacy of the decommissioning funding plan for RSNGS, (2) the probability of a loss of offsite power event at RSNGS, and (3) the adequacy of the staff er,vironmental review of decommissioning of RSNGS. CLI-93-03, slip op. at 32-33.

The staff incorporates by reference the ASLB determination in LBP-92-23 and the Commission decision in CLI-93-03 on all contentions and subjects, except those listed above.

. EC0, in its letter of March 22, 1993, filed amended contentions on the decommissioning funding plan. On April 1,1993, the licensee filed its response to these amended decommissioning funding plan contentions.

On April-12, 1993, the NRC staff filed its response to the amended decommissioning funding plan contentions.

In section 9.2 of this safety evaluation the staff addressed the adequacy of the RSNGS decommissioning funding plan in accordance with 10 CFR Part 50.82(a).

Also, in this section the staff addressed the earlier decommissioning funding comments submitted by ECO in its letter of June 15, 1992. The acceptability of the ECO contentions on the RSNGS decommissioning funding plan is now before the ASLB.

ECO, in its letter of April 1,1993, filed amended contentions on the probability of loss of offsite power. On April 13, 1993, the licensee filed its response to the amended loss of offsite power contentions.

On April 21, 1993, the staff filed its response to the same amended loss of offsite power contentions.

In section 8.0 of this safety evaluation and section 5.0 of the staff environmental assessment of decommissioning plan for RSNGS [Ref.10] the staff addressed the loss of offsite power issue and its impact on the health and safety of the public.

Acceptability of the ECO loss of offsite power contentions is now before the ASLB.

With regard to other contentions, the Commission in CLI-93-03 ordered, "ECO is to file any contentions in accordance with 10 CFR Section 2.714(b)(2)(iii),

regarding staff's environmental review, or other late-filed contentions, with the Licensing Board within 14 days after service of staff's environmental assessment or other documents reflecting staff's environmental review." The staff environmental assessment has been prepared and published.

Any contentions submitted by ECO will be dealt with in the adjudicatory process to the extent they raise substantial and pertinent issues that should be considered in relation to the RSNGS decommissioning plan. This matter is before the ASLB..

11.0 CONCLUSION

S The NRC concludes on the basis of the proceeding discussions, that (1) there is reasonable assurance that the health and safety of the public will not be 1

endangered by the decommissioning option selected by the licensee, (2) the major activities and tasks that take place during the decommissioning will be conducted in compliance with Commission regulations, and (3) the issuance of an order will not be inimical to the common defense and security or to the health and safety of the public.

. ~,,,,. - - -,.

7 y--

_.--.ro r

7..-,.-.

12.0 REFERENCES

(1)

Rancho Seco Nuclear Generating Station Unit' No.1, " Updated Safety Analysis Report."

(2)

U.S. Atomic Energy Commission, " Safety Evaluation by the Directorate of Iicensino. U.S. Atomic Energy Commission, in the Matter of Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station, Unit 1,"

Docket No. 50-312, June 1973.

(3)

" Rancho Seco Nuclear Generating Station Proposed Decommissioning Plan,"

(PDP) May 1991.

(4)

Rancho Seco Nuclear Generating Station, " Plan for Ultimate Disposition of the Facility" (PUDF), July 1990.

(5)

U.S. Nuclear Regulatory. Commission, " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," NUREG-0586, August 1988.

(6)

Sacramento Municipal Utility District, " Response to the Request for Additional Information in Support of the Rancho Seco Decommissioning Plan and Associated Environmental Report," April 15, 1992 (SMUD's responses to NRC questions).

(7)

Sacramento Municipal Utility District, "The Response to the Request for Additional Information in Support of the Rancho Seco Decommissioning Plan,"

August 6, 1992.

(8) Sacramento Municipal Utility District, " Response to the Request for

~

Additional Information in Support of the Rancho Seco Decommissioning Plan,"

August 31, 1992.

(9) Sacramento Municipal Utility District, " Supplement to the Applicant's Environmental Report - Post Operating Stage Rancho Seco Nuclear Generating Station," October 21, 1991.

(10)

U.S. Nuclear Regulatory Commission, " Environmental Assessment of the Rancho Seco Nuclear Generating Station, Unit-No. 1," February 1993.

(11)

Letter, from D. R. Keuter, Sacramento Municipal Utility District, to S. Weiss, U.S. Nuclear Regulatory Commission, May 20,-1991.

(12)

U.S. Nuclear Regulatory Commission, " Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station,"

NUREG/CR-0130, Volume 1, June 1978.

(13)

U.S. Nuclear Regulatory Commission, " Termination of Operating Licenses for i

Nuclear Reactors," Regulatory Guide 1.86, June 1974.

. (14) Amendment No. 119 to Facility Operating License No. DPR-54, Rancho Seco Nuclear Generating Station, " Permanently Defueled Technical Specifications," March 19, 1992.

(15)

" Standard Format and Content for Decommissioning Plan for Nuclear Reactors," Draft Regulatory Guide DG-1005, September 1989.

(16) American National Standards Institute (ANSI) Committee N18, Design Criteria for Nuclear Power Plants, " Selection and Training of Nuclear Power Plant Personnel," ANSI N18.1-1971.

(17) American National Standards Institute (ANSI), " Selection, Qualifications and Training of Personnel for Nuclear Power Plants," ANSI /ANS 3.1-1981.

(18)

U.S. Nuclear Regulatory Commission, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." NUREG-0800, June 1987.

(19)

Letter from S. W. Brown, Nuclear Regulatory Commission, to J. R. Shetler, Sacramento Municipal Utility District, dated March 19, 1992.

(20) Sacramento Municipal Utility District, " Radiological Characterization Plan for the Rancho Seco Nuclear Power Generating Station, April 1992 (Preliminary).

(21) Oak Ridge National Laboratory, "ANISN-W - Multi -Group One Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering," Oak Ridge National Laboratory Radiation Shielding Information Center, February 1986.

i (22) Oak Ridge National Laboratory, "0RIGEN2 - A Revised and Updated Version of the Oak Ridge Isotopic Generation and Depletion Code," Oak Ridge National Laboratory Radiation Shielding Information Center, October 1987.

(23)

Pacific Northwest Lab, " Residual Radionuclide Distribution and Inventory at Rancho Seco Nuclear Generating Station," PNL-5146, June 1984.

(24)

Lawrence Livermore Laboratory, " Environmental Radiological Studies.in 1989 Near the Rancho Seco Nuclear Power Generating Station," UCRL-ID-106111, November 1990.

(25)

U.S. Nuclear Regulatory Commission, " Calculation of Annual Doses -to Man from Routine Releases of Reactor Effluent for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Regulatory Guide 1.109, October 1977.

(26)

U.S. Nuclear Regulatory Commission, "Information Relevant to Ensuring That Occupational Radiation Exposures at Nuclear Power Stations Will Be As low As Is Reasonably Achievable," Regulatory Guide 8.8, Revision 3, June 1978.

(27)

U.S. Nuclear Regulatory Commission, " Operating Philosophy for Maintaining Occupational Radiation Exposure As Low As Is Reasonably Achievable,"

Regulatury Guide 8.10, Revision 1-R, September 1975.

?

i

. (28)

U.S. Nuclear Regulatory Commission, " Control of Radioactive Contaminated Materials," NRC IE Circular 81-07.

(29)

U.S. Nuclear Regulatory Commission, " Survey of Waste Before Disposal from Nuclear Reactor Facilities," NRC Information Notice 89-92.

(30)

Federal Reaister notice, June 27, 1988, Volume 53, Number 123, 24018 -

44056.

(31)

U.S. Nuclear Regulatory Commission, " Technical Position on Waste Form,"

Branch Technical Position, January 1991.

(32)

U.S. Environmental Protection Agency, "A Manual of Protective Actions for Nuclear Incidents," EPA 520/1-75-001.

(33)

"1992 Means Building Construction Cost Data."

(34)

" Dodge Manual for Building Construction Cost Data," 1984.

(35) Sacramento Municipal Utility District, " Decommissioning Plan Revisions,"

April 7, 1993.

(36)

Sacramento Municipal Utility District, " Decommissioning Plan Revisions Errata," April 19, 1993.

(37)

Sacramento Municipal Utility District, " Proposed Amendment No. 182, Revision 3, the Permanently Defueled Technical Specifications,"

November 19, 1991.

(38) Sacramento Municipal Utility District, " Revision to Permanently Defueled-Technical Specification Bases," September 23, 1992.

(39)

Sacramento Municipal Utility District, " Clarification of_ the Permanently Defueled Technical Specification Loop and SFP Decay Heat Analyses,"

Ap,'il 1, 1993.

(40)

" Final Environmental Statement related to the operation of Rancho.Seco Nuclear Generating Station Unit 1, Sacramento Municipal Utility District, Docket No. 50-312," U.S. Atomic Energy Commission, March 1973.

Principal Contributor:

Larry Bell a

Date: June 16, 1993

s DECOMMISSIONING SAFETY EVALUATION ABBREVIATIONS AGM ASSISTANT GENERAL MANAGER ALARA AS LOW AS REASONABLY ACHIEVABLE ANSI AMERICAN NATIONAL STANDARDS INSTITUTE BWST BORATED WATER STORAGE TANK CBS CONTAINMENT BUILDING SPRAY CFH CERTIFIED FUEL HANDLER CFR CODE OF FEDERAL REGULATIONS CFS CORE FLOOD SYSTEM CM CLOSURE MANAGER COTP CALIFORNIA /0REGON TRANSMISSION PROJECT CPM COUNTS PER MINUTE CRDM CONTROL R0D DRIVE MECHANISM DHS DECAY HEAT SYSTEM DOC DECOMMISSIONING OPERATIONS CONTRACTOR DPM DISINTEGRATIONS PER MINUTE ECO ENVIRONMENTAL AND RESOURCES CONSERVATION ORGANIZATION EFPD EFFECTIVE-FULL-POWER DAY EPA ENVIRONMENTAL PROTECTION AGENCY FR FEDERAL REGISTER FSV FORT SAINT VRAIN GDC GENERAL DESIGN CRITERIA GET GENERAL EMPLOYEE TRAINING HEPA HIGH-EFFICIENCY PARTICULATE AIR HVAC HEATING, VENTILATION, AND AIR CONDITIONING ICI INCORE INSTRUMENTATION IE OFFICE OF INSPECTION AND ENFORCEMENT 10SB INTERIM ONSITE STORAGE BUILDING ISFSI INDEPENDENT SPENT FUEL STORAGE INSTALLATION LOOP LOSS OF ALL UNIT AC POWER MESC MULTI-ELEMENT SEALED CANISTER MPC MAXIMUM PERMISSIBLE CONCENTRATION NRC NUCLEAR REGULATORY COMMISSION NSRAC NUCLEAR SAFETY REVIEW AND AUDIT COMMITTEE i

NSSS NUCLEAR STEAM SUPPLY SYSTEM j

ODCM "0FFSITE DOSE CALCULATION MANUAL" 0TSG ONCE-THROUGH STEAM GENERATOR PAG PROTECTIVE ACTION GUIDELINE-PDP

" PROPOSED DECOMMISSIONING PLAN" PDTS PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS I

PLS PURIFICATION AND LETDOWN SYSTEM PNL PACIFIC NORTHWEST LAB POL POSSESSION ONLY LICENSE i

PRT PRESSURIZER RELIEF TANK PUDF

" PLAN FOR ULTIMATE DISPOSITION OF THE FACILITY" PWR PRESSURIZED-WATER REACTOR QA QUALITY ASSURANCE QAP QUALITY ASSURANCE PROGRAM

" RANCHO SECO DEFUELED SAFETY ANALYSIS REPORT" l

RSNGS RANCHO SECO NUCLEAR GENERATING STATION RSSQM RANCHO SECO SAFSTOR QUALITY MANUAL RWP RADIATION WORK PERMIT SER SAFETY EVALUATION REPORT SFP SPENT FUEL P00L SIM SAFETY INJECTION AND MAKEUP SMUD SACRAMENTO MUNICIPAL UTILITY DISTRICT SPC SPENT FUEL POOL COOLING TLD THERM 0 LUMINESCENCE DOSIMETER USAR

" UPDATED SAFETY ANALYSIS REPORT" 8

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