ML20204C071

Safety Evaluation Supporting Amend 102 to License NPF-58

ML20204C071
Person / Time
Site:	Perry
Issue date:	03/11/1999
From:	NRC (Affiliation Not Assigned)
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ML20204C058	List: ML20204C052 ML20204C071
References
NUDOCS 9903220369
Download: ML20204C071 (14)

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g UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3000H001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.102 TO FACILITY OPERATING LICENSE NO. NPF-58 FIRSTENERGY NUCLEAR OPERATING COMPANY PERRY NUCLEAR POWER PLANT. UNIT 1 QQCKET NO. 50-440

1.0 INTRODUCTION

By letter dated November 2,1995, as supplemented by submittal dated January 7,1999, FirstEnergy Nuclear Operating Company (the licensee, formerly The Cleveland Electric illuminating Company and Centerior Service Company) proposed changes to the Perry Nuclear Power Plant, Unit 1 (PNPP) Technical Specifications (TSs) during the handling of irradiated fuel in the Primary Containment and Fuel Handling Building, and selected specifications associated with CORE ALTERATIONS. The purpose is to establish a point where OPERABILITY of those systems typically used to mitigate the consequences of a fuel handling accident (FHA) is no longer required to meet the Standard Review Plan guidance on offsite dose effects (i.e., less than 25% of 10 CFR Part 100 limits). Specifically, the proposal identifies that only "recently"

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irradiated fuel contains sufficient fission products to require OPERABILITY of accident mitigation features to meet the accident analysis assumptions. Therefore, the APPLICABILITY j

requirements for the associated mitigation features are revised.

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The licensee has performed a fuel handling accident (FHA) dose analysis which takes credit for a radioactive decay period that is longer than the 24-hour period originally assumed. Given this l

longer decay period, the licensee proposed changes to redefine the TS requirements by relaxing Primary Containment, Secondary Containment, and Fuel Handling Building integrity requirements and relaxing requirements for those engineered safety feature systems originally relied upon to mitigate a FHA. To implement the above concepts, these TSs will only apply if fuel has been "recently irradiated." The term "recently irradiated"is a cycle specific number and represents the decay period for the reduction in radionuclide inventory available for release in j

the event of an FHA. For the upcoming refueling outage, the licensee has determined that the appropriate decay period will be 7 days. In summary, once the reactor has been shut down for a minimum of 7 days, the licensee has demonstrated that the FHA reanalysis (that does not rely on either building integrity or the FHA mitigating systems) will not exceed offsite dose limitations.

The TS Bases will be revised to provide a cycle-specific definition of "recently irradiated" fuel.

The supplemental information contained clarifying information and did not change the initial no significant hazards consideration determination and did not expand the scope of the original application.

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2.0 BACKGROUND

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2.1 Reaulatorv Reauirements for Shutdown Ooeration Historic development of regulatory requirements for nuclear power plant operation was based on the premise that most potential risk was due to operation at power and, consequently, protection of the public could be ensured by designs and operations that conservatively

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bounded all conditions by achieving defense-in-depth for power operation. Fuel movement was recognized as a departure from this concept since there was no corresponding power op6. ration configuration, and this was judged as an area where additional regulatory protection was necessary. This is reflected in technical specifications where there are many containment requirements during power operation, but during Cold Shutdown and Refueling Modes few requirements apply outside of fuel handling and related operations.

During the late 1980s and early 1990s, the staff and industry realized that significant risk reductions could be achieved during shutdown operation. The staff responded with a i

rulemaking effort and industry implemented voluntary initiatives to realize risk improvements. In im,ognition of these efforts, work to improve technical specifications was concentrated on power operation specifications, with the intention to address shutdown once a rule was in place. As a result, shutdown technical specifications are not consistent with the reduction in risk that would be achieved from increased emphasis on technical specifications.

l In regard to shutdown operation, on July 30,1997, the staff credited the effectiveness of industry's voluntary actions in well-operated plants by informing the Commission that such voluntary ".. initiatives have been successful in achieving the acceptable level of risk that now exists at U.S. nuclear power plants" and "The practical effect of rule implementation is, therefore, not to raise the current level of safety, but rather to ensure that at least the current level of safety will be maintained."' On December 11,1997, the Commission decided not to issue a shutdown rule for comment. Instead, the Commission instructed the staff to "... continue to monitor licensee performance, through inspections and other means, in the area of shutdown operations j

to ensure that the current level of safety is maintained."2 The major component of the Commission's decision to not issue a shutdown rule was the effective voluntary actions in place in the well-run nuclear power plants, and the expectation that those or equally effective actions l

would continue.

One aspect of enhanced understanding of shutdown operation is an understanding that the risk due to potential fuel handling accidents, particularly if the decay heat generation rate is low, is almost nil, whereas the risk due to many other shutdown operations is comparable to, and sometimes exceeds, the risk during power operation. Yet, there are more restrictive technical 1

' Issuance for Public Comment of Proposed Rulemaking Package for Shutdown and Fuel 4

Storage Pool Operation," SECY-97-168, July n 1997.

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' Staff Requireme.e SECY-97-168 -Issuance for Public Comment of Proposed Rulemaking Package for Shutdown and Fuel Storage Pool Operation," Staff Requirernents Memorandum, December 11,1997.

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specifications for fuel handling compared to other aspects of shutdown operation. With respect to containment during the Cold Shutdown and Refueling Modes, the only requirement applies to fuel movement and related operations; there are no other containment requirements. The Perry licensee has recognized th;s paradox, and is proposing to relax technical specification requirements during fuel handling when an appropriately low decay heat generation rate has been achieved while committing to continue to ensure an available containment during Cold Shutdown and Refueling Mode operation via administrative procedures. The licensee states that this technical specification relaxation will permit the optimization of outages to achieve an overall risk reduction while also reducing outage time and cost. A significant contributor to this risk reduction is the ability to postpone operations early in the outage that, from a practical standpoint to achieve a short outage time, must be performed soon after shutdown when there is no technical specification requirement for a closed containment. The requested amendment will allow some of these operations to be accomplished later, when the reactor vessel is open and covered by 23 feet of water - when the risk of a severe core damage accident is almost nil.

The trade-offs between the requested technical specification relaxation during fuel handling and the voluntary actions to achieve containment closure during Cold Shutdown and Refueling Mode operations, with a corresponding reduction in risk, are basic to the staff's approval of the licensee's request.

2.2 Oriainal Reauirements/ Licensee's Proposal The Perry licensee has implemented NUREG-1434, Revision 1, "BWR-6 Improved Standard Technical Specifications." These TSs have a number of operational restrictions during shutdown conditions. The shutdown conditions requiring TS OPERABILITY are captured in the APPLICABILITY statements of the TS. Tne standard wording of the APPLICABILITY statements during shutdown are as follows:

During movement of irradiated fuel assemblies in the primary containment, During CORE ALTERATIONS, During operations with a potential for draining the reactor vessel (OPDRVs).

Structures such as the Primary Containment, Secondary Containment, and Fuel Building must be OPERABLE during the above conditions. Similarly, systems related to performing core alterations must also be OPERABLE during the above conditions. However, outside of the above conditions, OPERABILITY of the Primary Containment, Secondary Containment, Fuel Building, and systems related to performing core alterations are not required.

During refueling outages, movement of large equipment into or out of Primary Containment or the Fuel Building (such as chemical-decontamination equipment, inservice examination / test equipment, or large component parts that require repair) must either be completed prior to establishing OPERABILITY or delayed until after OPERABILITY is required. Real dollar losses are incurred due to the inability of specialized contractors to perform their designated activities due to delays in performance of critical path activities. Also, productivity losses occur when personnel are involved in multiple evolutions of establishing, maintaining, and releasing OPERABILITY. These factors, coupled with the increased flexibility for scheduling testing and maintenance activities on containment valves and instrumentation, can result in significant accrued cost reductions and productivity enhancements over the remaining operating life of the

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. plant, allowing outage resources to be directed to other activities, which ultimately will result in improvements in plant maintenance, operations, and overall safety.

The Perry licensee has proposed to relax TS requirements during shutdown conditions. The premise is to take credit for the normal decay of irradiated fuel, reanalyze the design basis accident during shutdown conditions (i.e., the FHA), and thus conclude that neither building integrity nor the FHA mitigating systems are required to be OPERABLE during shutdown conditions.

j On many plant dockets, including PNPP, the NRC has determined that the FHA is acceptable when conservatively calculated dose analyses result in doses which remain less than 25% of 10 CFR Part 100 guidelines. This is also reflected in Standard Review Plan 15.7.4,

" Radiological Consequences of Fuel Handling Accidents." Typically for Boiling Water Reactors (BWR's), these types of dose analyses show that fuel handling is acceptable to begin once 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has passed after entry into a plant shutdown. Perry has installed filtration capabilities in the ventilation system for the fuel handling area and their current analyses take credit for the filtration in reducing the doses when performing such dose calculations.

The attemative approach being proposed is to take credit for the normal decay of irradiated fuel rather than crediting the active mitigative systems (e.g., ventilation and filtration systems). Since radioactive decay is a natural phenomenon, it has a reliability of 100 percent in reducing the radiological release from the fuel bundles. In addition, the water level that covers the fuel bundles is another natural method that provides an adequate barrier to a significant radiological release and this defense-in-depth method will continue to be enforced by Technical Specification controls (i.e., TS 3.9.6, " Reactor Pressure Vessel (RPV) Water Level - Irradiated Fuel," requires that RPV water level be 222 feet 9 inches above the top of the RPV flange).

By letter dated November 2,1995, the licensee provided a revised offsite dose calculation showing that the consequences of a FHA would remain less than 25% of 10 CFR Part 100 guidelines discussed above, once the fuel had undergone radioactive decay for several days.

The length of this "several day" period is determined by a plant-specific dose calculation. The analysis took no credit for the primary containment, the fuel handling building, or the installed ventilation systems (including their filtration capabilities) after this extended period of decay.

The submittal proposed that the NRC should permit core alteration / fuel handling activities to occur after this period of radioactive decay, without requiring TS controls over building integrity and ventilation system / filtration operability. The period of decay that was assumed for Perry is 7 days. Thus,7 days following reactor shutdown, the licensee's analyses show that due to the natural decay of irradiated fuel, the off-site dose resulting from the FHA will not exceed 25% of 10 CFR Part 100 even if credit is not taken for building integrity or FHA mitigating systems.

The licensee proposed large-scale relaxations to the TS by revising the APPLICABILITY statements for shutdown conditions for structures (e.g., Primary Containment, Secondary Containment, Fuel Handling Building) and systems previously used to mitigate the consequences of a FHA. The APPLICABILITY statements were to be revised as follows:

6 1 i During movement of recently irradiated fuel assemblies in the primary containment or fuel handling building, Du;ir,i, OOil ALTEMTlON6; During operations with a potential for draining the reactor vessel (OPDRVs).

i In order to implement the above APPLICABILITY statements, the Limiting Condition for Operation (LCOs) for INTEGRITY and for the selected ESF systems need only apply if fuel that has recently been in the critical reactor core (i.e., "recently irradiated fuel") is handled during the first several days of an outage (prior to completion of the longer decay period). The TS Bases will be revised to identify "recently irradiated fuel" as fuel that has (,ccupied part of a critical reactor core within the previous 7 days.

The deletion of the CORE ALTERATIONS term is justified since a FHA is the only event during j

CORE ALTERATIONS that is postulated to result in fuel damage and radiological release, and such FHAs will be fully enveloped by the proposed APPLICABILITY.

in addition to the above changes to the APPLICABILITY statements, the licensee proposed j

numerous corresponding changes to the ACTION statements, such as elimination of references i

to CORE ALTERATIONS and the insertion of 'recently irradiated fuel" when referring to the movement ofirradiatad fuel.

1 The proposed changes do not impact TS requirements for systems r.oeded to prevent or j

mitigate CORE ALTERATION events other than the FHA. They also do not change the i

requirements for systems needed to mitigate potential vessel c:raindown events, systems j

needed for decay heat removal, or the requirements to ma;ntain high water levels over irradiated fuel.

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b O The licensee proposed changes to the following TS:

SPECIFICATION TITLE TS NUMBER Primary Containment and Drywell isolation Instrumentation 3.3.6.1 Control Room Emergency Recirculation (CRER) System Instrumentation 3.3.7.1 Primary Containment Air Locks 3.6.1.2 Primary Containment Isolation Valves (PCIVs) 3.6.1.3 Primary Containment - Shutdown 3.6.1.10 Containment Vacuum Breakers 3.6.1.11 Containment Humidity Control 3.6.1.12 Secondary Containment 3.6.4.1 Secondary Containment isolation Valves (SCIVs) 3.6.4.2 Annulus Exhaust Gas Treatment (AEGT) System 3.6.4.3 Control Room Emergency Recirculation (CRER) System 3.7.3 Control Room HVAC System 3.7.4 Fusi Handling Building 3.7.8 Fuel Handling Evilding Ventilation Exhaust System 3.7.9 AC Sources - Shutdown

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Sources rMdown*

3.8.5 F

1 j Distribution Systems - Shutdown

• 3.8.8

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• TSs for electrical systems do not have CORE ALTERATIONS in their APPLICABILITY statements. Since these TSs also havs MODES 4 and 5 in their APPLICABILITY statements, OPERABILITY will always be applicable provided that fuel remains in the reactor vessel.

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2.3 Limited Staff Anoroval for Containment Personnel Airlocks By letter Ated February 2,1996, the staff issued Amendment No. 80 to the Perry operating license, in;s was a partial approval of the licensee's submittal and only approved opening of the containment personnel air locks (TS 3.6.1.2 ) during shutdown conditions. The staff's review includsd a reanalysis of the FHA under the premise that the reactor had been shut down for a minimum of seven days. The staff assumed an instantaneous puff release of noble gases and 4

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i t l rad loiodines from the gap and plenum of the broken fuel rods. These gas bubbles were assumed to pass through at least 22 feet 9 inches of water covering the fuel prior ro reaching i

the containment atmosphere. All airborne activity reaching the containment was assumed to j

exhaust from the plant vent to the environment within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The staff approval was i

contingent upon licensee procedures to close one personnel airlock door quicitly in the event of an accident in order to establish containment integrity.

While approving extended opening of the containment personnel airlocks during plant shutdown, the staff deferred taking action on the licensee's larger request since, as discussed in Section 2.1 above, ongoing activities related to the shutdown rule were anticipated to provide additional j

s'.aff guidance on TSs during plant chutdown.

i 2.4 Industry Iritiative On September 8,1998, the BWR/6 owners met with the NRC staff to readdress the FHA reanalysis. Since the original proposal of November 1995, the staff had granted limited approval i

to permit opening of the containment personnel airlock doors for most plants but had deferred taking additional actions. Modeled after Perry's original plant-specific proposal, the Nuclear Energy institute (NEI) proposed generic changes to the improved Standard Technical Specifications via Technical Specification Task Force Traveler 51 for all four owners groups on March 25,1996. During the public meeting of September 8,1998, it was agreed that the Perry facility would be treated as the lead plant for this activity.

3.0 EVALUATION The staff's review focused on the following four areas:

5 (1) Dose Calculations - Control room and offsite dose consequences must be within acceptable regulatory limits without taking credit for the integrity of the Primary Containment, Secondary Containment, and Fuel Handling Building as well as the FHA mitigating systems.

(2) Administratve Controls - Shutdown safety controls must address 1) procedures to assess the impact of removing systems from service during shutdown conditions,

2) the ability to implement prompt methods to close both the Primary Containment and the Fuel Hanoung Building in the event of a FHA, and 3) controls to avoid unmonitored releases.

(3) Risk Significance - The licensee's risk-related discussion needs to support the proposed TS changes.

(4) Shutdown Operations - The licensee's proposed amendment should be consistent with the Commission's December 11,1997, instructions to the staff.

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3.1 FHA Reanalysis The staff reviewed the licensee's justification for allowing relaxation of the Primary Containment L

and Fuel Handling Building integrity requirements during fuel handling activities not involving i

l recently irradiated fuel. The licensee defines "recently irradiated fuel" as fuel that has occupied i

l part of a critical reactor core within the previous 7 days. As part of this review, the staff reviewed the licensee's reanalysis of the FHA.

i The licensee's revised FHA analysis assumed an instantaneous release to the environment with no holdup or treatment (i.e., no credit was taken for building integrity or FHA mitigating systems).

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The offsite dose consequence results were previously submitted to the NRC by letter dated March 16,1990, in suppoit of a request for approval of TS changes to allow opening of up to six 3/4-inch vent and drain line pathways during refueling activities, in review of that request, the staff confirmed the licensee offsite dose results and found they met 10 CFR Part 100 acceptance criteria (see the NRC Safety Evaluation for PNPP Operating License NPF-58, Amendment No. 35, dated September 28,1990.) No new or different offsite dose analyses l

were performed for this current request. The staff has determined that the previously submitted l

offsite dose analysis and NRC staff review are applicable to this request.

By letter dated January 7,1999, the licensee submitted the control room dose analysis for the l

fuel handling accident. The analysis assumed a 7-day decay period and took no credit for l

Containment or Fuel Handling Building integrity and the FHA mitigating systems. The licensee's l

analysis also assumed that the Control Room Emergency Recirculation System does not initiate, while the normal control room ventilation system continues to run and draw in unfiltered j

outside air at 6,600 cfm. The licensee's dose results were as follows:

. Licensee Calculated Control Room Doces (0-30 days)

Calculated GDC 19 Acceptance Criteria Whole Body Gamma Dose (rem) 0.0125 5

Inhalation Dose (rem) 27.3 30 l

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- The staff performed an independent control room radiological dose analysis using the same assumptions with regard to control room ventilation and containment integrity. These j

assumptions are listed below.

i FHA Control Room Dose Assumptions Used by NRC Staff No credit for containment integrity (direct release of radioactivity to the environment) i Normal control room ventilation No filters on control room ventilation intake Power Level 3758MWth Radial peaking factor 1.5 l

Radioactive decay period 7 days Fuel Rods damaged 98 (1 assembly)

Total fuel rods in core 59,400 1

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lodine gap activity fractions Organic 0.0025 Inorganic 0.9975 Control Room X/Qs i

0-P hours 3.5E-04 8-24 2.1 E-04 14 days 1.1 E-04 4-30 days 5.75E-05 The results of the staff calculation were:

NRC Staff Calculated Control Room Doses (0-30 days)

Calculated GDC 19 Acceptance Criteria Whole Body Gamma Dose (rem)

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inhalation Dose (rem) 23.1 30 1

Tne licensee's control room dose results are comparable to staff calculated doses and meet the 2

acceptance criteria given in 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 19.

The staff finds the licensee's control room radiological dose analysis and results to be conservative and acceptable. Taking into consideration the previously reviewed FHA offsite 2

dose calculations, the staff finds the proposed changes acceptable with regard to radiological

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consequences.

e 3.2 Shutdown Safety Controls The grea of reWew under shutdown safety controls focused on 1) procedures to assess the impact of removing systems from service during shutdown conditions, 2) the ability to implement prompt methods to close both the Primary Containment and the Fuel Handling Building in the event of a FHA, and 3) contrals to avoid unmonitored releases.

In the licensee's submittal of November 2,1995, it referenced Section 4.5 of NUMARC 91-06,

" Guidelines for Industry Actions to Assess Shutdown Management." NUMARC 91-06 focused on events involving loss of decay heat removal and addressed the ability to promptly restore containment integrity. It identified that the time to effect closure should be consistent with plant conditions (e.g., reactor coolant inventory and decay heat load). In this regard, the licensee developed administrative controls for the closure of the Primary Containment and the Fuel i

Handling Building, which were based on the recommendations of NUMARC 91-06 Section 4.5.

Subsequent to the development of NUMARC 91-06, the staff completed its activities associated with the Shudown Rulemaking. The Shutdown Rulamsking did not result in any additional TSs during shutdown condit!ons. With regard to NP.C concems over removal of significant systems from service during p. ant shutdowns, the Commission directed the staff to address these concems by placing new limitations in the maintenance rule (i.e.,10 CFR 50.65, " Requirements for monitoring the effectiveness of maintenance at nuclear power plants"). The proposed change to 10 CFR 50.65 would require licensees to assess the impact on shutdown safety before removing equipment from service for maintenance.

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! The industry, through the Nuclear Energy Institute (NEI), has been developing guidance to implement this Commission directive. A revised draft of NUMARC 93-01, " Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants was submitted to the NRC on July 1,1998. While NUMARC 91-06 only focused on selected shutdown operations,

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NUMARC 93-01 addressed a broad scope of activities during shutdown conditions.

In the draft NUMARC 93-01 guideline, Section 11.2.6, " Safety Assessment for Removal of Equipment from Service During Shutdown Conditions," under the subheading of " Containment -

Primary (PWR)/ Secondary (BWR)", the following guidance is provided.

" ibrplants which obtain amendments to modify Technical Specification requirements on pnmary 1

or secondary containment operability and ventilation system operability during fuel handling or core alterations, the following guidelines should be included in the assessment of systems removed from service:

During fuel handling / core alterations, ventilation system and radiation monitor availability (as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly.

The basis of the Technical Slmcification operability amendment is the reduction in doses due to such decay. The goal of mt%aining ventilation system and radiation monitor availabilityis to reduce doses even further below that provided by the natural decay.

A single normal or contingency method to promptly close primary or secondary containment penetrations shou!d be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure."

The purpose of the " prompt methods" mentioned above is to enable ventilation systems to draw the release from a postulated fut inandling accident in the proper direction such that it can be treated and monitored.

The draft NUMARC 93-01 guidance is built upon two basic premises: avoiding unmonitored releases and using available (although not necessarily " Technical Specification OPERABLE")

filtration capabilities to reduce doses below those achieved from the decay of the source term and the scrubbing of the water. Until such time that NUMARC 93-01 is endorsed as a formal industry position, the Perry licensee has committed to the above draft wording for controlling the removal from service of systems, structures and components that are currently required by TSs during periods of core alteration / fuel handling.

In response to its commitment to NUMARC 93-01, the licensee updated its administrative i

controls for Primary Containment / Fuel Handling Building closure. These closure controls are in effect whenever the Primary Containment or Fuel Handling Building is open and are not limited to fuel handling operations. Areas addressed in these administrative controls include the following:

• Equipment necessary to implement containment clesure shall be appropriately staged prior to maintaining airlock doors open.

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• Hoses and cables running through any open penetration, airlock, or hatch should be tagged to facilitate rapid removal in the event that containment closure is required.

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• One door in each airlock is capable of being closed and the airlock door is not blocked in l

such a way that it cannot be expeditiously closed. In addition, personnel are designated with j

the responsibility for expeditious closure of airlock doors.

. Major disassembly of containment boundary valves, except those valves 3/4-inch or less, should only be performed on one valve at a time with administrative controls established on the opposite boundary valve. If conditions require working both containment isolation valves in parallel, closure devices shall be fabricated and staged at the work area.

l Major ventilation and air conditioning systems, including radiation release monitoring, shall l

be available. Such HVAC systems include the Control Room, Fuel Handling Building, l

Annulus Exhaust Gas Treatment, and Primary Containment Building.

Compensatory actions for prompt closure of the Primary Containment / Fuel Handling Building will give priority to the following: Containment airlock doors, Fuel Handling Building roll-up door and personnel doors, containment ventilation boundary valves, feedwater isolation valves, main steam isolation valves, annulus exhaust gas treatment system, and l

containment vacuum relief valves.

• Personnel responsible for Primary Containment / Fuel Handling Building closure shall be trained and knowledgeable in using procedures for re-establishing building integrity.

The staff has reviewed the Perry administrative procedures on closure and concludes that they provide reasonable and adequate controls to achieve Primary Containment / Fuel Handling Building closure.8 in accordance with regulatory requirements, the licensee must develop procedures to maintain control of radioactive effluents and to maintain doses to members of the public from radioactive effluents as low as reasonably achievable. The licensee's program for these requirements are described in TS 5.5.4, " Radioactive Effluent Controls Program." The staff notes that the licensee's Radioactive Effluent Controls Program is not impacted by these proposed TS changes and, therefore, a situation will not occur that could result in an unmonitored release.

The staff considers the licensee's described administrative controls as an adequate means to control monitoring and filtration of any releases that might occur froin a FHA and to be consistent with the Commission's De. ember 11,1997, instructions to the staff. Therefore, the staff concludes that the licensee's shutdown safety controls for building integrity and ventilation / filtration systems is an acceptable means of supporting the proposed TS changes.

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• Reasonable and adequate controls means an integral barrier or controlled filtration can be provided in time to control a significant release of radioactive material end to achieve an adequate means to control monitoring of releases.

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. 1 3.3 Risk Evaluation There have been several occurrences in the history of the nuclear power industry in which fuel bundles have actually been dropped in the course of fuel handling activities. In each of these Instances, the actual releases from the fuel have been minimal or nonexistent (reference NSAC/129 and other subsequent plant operating event reports). This has shown that the assumptions utilized in the radiological dose calculations for a fuel handling accident are quite conservative.

An exsmination of the significance of Fuel Handling Accidents was examined as part of a Grand Gulf shutdown risk study (reference NRC Meeting Summary of September 9,1998, " Meeting To Discuss The Planned Joint Proposals On Containment Requirements To Mitigate Fuel Handling Accidents During Refueling" with several BWR/6 plants), insights from this study show that due to the much lower potential releases from a Fuel Handling Accident than from a coie damage accident (approxie.ately 100 Curies as compared to 3 X10' Curies) the risk from a Fuel Handling Accident is very low, and is 3 orders of magnitude below the risk associated with a core damage event during shutdown.

The staff has reviewed the licensee's risk-informed discussion and supports the proposed license amendment for the following reasons:

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• Results of agency sponsored probabilistic risk assessment (PRA) studies for Grand Gulf, a l

plant of similar design to Perry, indicate that during shutdown, the potential for core damage is l

least when the reactor vessel head is off (thus alleviating concerns regarding overpressurization of shutdown cooling system components) and the vessel water level is raised (thereby providing more time for mitigation of accident initiating events). During refueling activities when fuel movement is taking place, TSs require a minimum water level of 22 feet 9 inches of water above the active core. This is the case of the plant operating state i

(POS) associated with fuel handling during refueling outages.

+ There are no TSs requiring containment integrity during shutdown other than the one involving fuel handling (even though the risk associated with some of these POSs is higher).

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Furthermore, no such TSs were propcsed to address core damage related concerns reised during the Shutdown Rulemaking process.

PNPP has outage management administrative controls in place for re-establishing containment closure consistent with plant conditions.

+ Increases in core damage frequency (CDF) and large early release frequency (LERF),

associated with the proposed change, would most likely be considerably less than 1E-6/yr and 1E-7/yr, respectively.

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l l-3.4 Summary The proposed TS changes redefine the fuel handling requirements in two areas, given the longer decay period from the time of reactor subcriticality:

Requirements associated with INTEGRITY for the Primary Containment and Fuel l

Handling Building are relaxed (since no credit is taken 'or these in the new analysis for I

mitigation of a FHA).

1 Requirements for selected engineered safety feature systems (those that are not credited in the new analysis for mitigation of a FHA).

The proposed changes do not impact TS requirements for systems needed to prevent or mitigate CORE ALTERATION events other than the FHA. They also do not change the requirements for systems needed to mitigate potential vessel draindown events, systems needed for decay heat removal, or the requirements to maintain high water levels over irradiated fuel.

As previously discussed in this evaluation, the staff finds the proposed TS changes acceptable because:

Fuel handling accidents are not risk-significant and have not merited individual TS controls.

Adequate defense in depth is maintained by the requirements for water level and the natural decay ofirrediated fuel.

The control room and offsite dose calculations meet the acceptance criterion without reliance on building integrity or FHA mitigating systems.

Administrative controls over shutdown safety are in effect that ensure containment closure, should it be need=>d, and to control monitoring and filtration of any releases that might occur from a FHA.

Risk-informed considerations support the licensee's proposed TS charges.

4.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Ohio State official was notified of the proposed issuance of the amendment. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

This amendment changes a requirement with respect to insWlation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement. The staff has determined that the amendment involves no significant l

o.i - increase in the amounts, and no significant change in tlie types, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (60 FR 62497). Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

6.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor (s):

Michelle Hart Nicholas Saltos i

Warren Lyon Douglas Pickett Date: March 11, 1999 l

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