ML20202D162

Safety Evaluation Supporting Amend 193 to License DPR-61

ML20202D162
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Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/30/1998
From:	NRC (Affiliation Not Assigned)
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References
NUDOCS 9902010323
Download: ML20202D162 (23)

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k UNITED STATES j NUCLEAR REGULATORY COMMISSION I

$ f WASHINGTON. D.C. 20$55 0001 f k .... / -

1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION l

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RELATED TO AMENDMENT iNO.193 TO FACILITY OPERATING LICENSE NO. DPR- 61 i

CONNECTICUT YANKEF ATOMIC POWER COMPANY  !

CONNECTICUT YANKFF ATOMIC POWER STATION i i

DOCKET NO. 50-213

1.0 INTRODUCTION

i By letter dated May 30,1997, identified as CY 97-006, as supplemented by letters dated  !

May 7,1998 (CY-98-071) and June 18,1998 (CY-98-108), Connecticut Yankee Atomic t Power Company (CYAPCo or the licensee) submitted a request for a change to the  ;

Connecticut Yankee Atomic Power Station (Connecticut Yankee), also known as the l Haddam Neck Plant (HNP), Facility Operating Uconse No. DPR-61. The May 7,1998, .!

Supplement relocated the provisions of Technical Specification 3/4.9.15. The June 18,  !

1998, supplement consisted of supporting technicalinformation. The supplements did not change the staff's initial proposed no significant hazard consideration determination or expand the scope of the original notice. The requested change would replace in their l entirety the existing Technical Specifications (TS) incorporated in the Connecticut Yankee i Facility Operating Ucense as Appendix A. In a second letter. dated May 30,1997, identified i as CY-97-024, the licensee requested approval to replace licensed operators with Certified i Fuel Handlers, which included changes to Section 6.0 of the TS. The changes in Section l 6.0 requested in the two letters were consolidated into Ucense Amendment 192, issued on 1 March 27,1998. The remaining changes in CY-97-036, as supplemented, are evaluated in this safety evaluation. Connecticut Yankee developed the revised Technical Specifications, titled Defueled Technical Specifications (DTS), to reflect the permanently shutdown and i defueled status of the plant. Changes are proposed in the definitions, limiting conditions for i operations, limiting conditions for operation, surveillance, design features, bases, and l administrative control sections.

2.0 DISCUSSION AND EVALUATION I

l On December 4,1996, the Connecticut Yankee Atomic Power Company (CYAPCo) Board of L , Directors decided to permanently cease further operation of the Connecticut Yankee Nuclear l- Plant. On December 5,1996, in accordance with 10 CFR 50.82(a)(1), CYAPCo provided to L

the U.S. Nuclear Regulatory Commission (NRC) certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel. The changes to the

! Facility Operating Ucense and TS proposed by the licensee reflect the limitations and requirements appropriate to the present configuration of the plant. The primary reason for the change was to simplify and to improve clarity by eliminating the large volume of non-applicable material in the current Operating Ucense and Technical Specifications, and obtain a clear and concise document for maintaining the plant in a permanently defueled i condition.

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l 2-Connecticut Yankee's current Technical Specifications (CTS) have been customized over the plant's life to meet the specific needs of the facility, and were revised in Amendment 125 on April 26,1990 to the format used in the Westinghouse Standard-format Technical Specifications. The Commission issued NUREG 1431, Standard Technical Specifications (STS), Westinghouse Plants, in September 1992, which was developed using the guidance and criteria contained in the Commission's Interim policy statement. STS were established as a model for developing improved TS for Westinghouse plants in general. The generic bases presented in NUREG-1431 provide information regarding the extent to which the STS present requirements that are necessary to protect the public health and safety. Although the Standard Technical Specifications were developed for operating plants, the STS also contain the latest approved NRC guidance on technical specifications for a plant that does not operate but that has irradiated fuelin the spent fuel pool. The safety functions related to safe maintenance and storage of irradiated fuel at an operating plant are similar to the corresponding safety function at a permanently shutdown plant.

In general, the proposed changes to the requirements in Connecticut Yankee's current Technical Specifications were either retained, deleted due to their inapplicability to a f acility with a defueled reactor, or relocated to appropriate licensee-controlled documents in accordance with existing NRC guidance. Technical Specification renuirements designateri for relocation were proposed to be relocated from the Technical Specifications to the referenced licensee-controlled documents (Technical Requirements Manual). The licensee added a specification for a spent fuel pool cooling and makeup monitoring program that was not in Connecticut Yankee's revised standard Technical Specifications. The licensee also added a specification to continue the applicability of the spent fuel pool (SFP) temperature limit to the permanently defueled condition. Therefore, the cumulative effect of the requirements in the Updated Final Safety Analysis Report (UFSAR), Operating License, proposed Technical Specifications, and other regulations is to continue to ensure safe maintenance and storage of irradiated fuel at the Connecticut Yankee facility.

BACKGROUND Section 182a of the Atomic Energy Act (the Act) requires that applicants for nuclear power plant operating licenses will state:

(Sluch technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization... of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS those items that are directly related to maintaining the integrity of the physical barriers designed to contain l

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i radioactivity." Statement of Consideration. " Technical Specifications for Facility Ucenses,

~ Safety Analysis Reports," 33 FR 18610 (December 17, 1968). Pursuant to 10 CFR 50.36,

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TS are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings and limiting control settings; (2) limiting conditions for  !

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- operation (LCOs); (3) surveillance requirements (SR); (4) design features; and l  ;

(5) administrative controls. However, the rule does not specify the particular requirements t

to be included in a plarit's TS. )

On July 22,1993, the Commission issued its Final Policy Statement, expressing the view '

that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36 (58 FR 39132). The Final Policy Statement described the safety benefits 3 of the improved STS, and gave guidance for evaluating the required scope of the TS and j defined the guidance criteria to be used in determining which of the LCOs and associated  ;

i surveillances.should remain in the TS. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be

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retained in the TS, it was adopting the qualitative standard enunciated by the Atomic Safety j

and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear plant),

, ALAB-531,9 NRC 263,273 (1979). There, the Appeal Board observed: ,

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l (T!bere is neither a statutory nor a regulatory requirement that every operational i

detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition

l of operation which is legally binding upon the licensee unless and until changed  !

with specific Commission approval. Rather, as best we can discern it, the  !

contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid ,

conditions or limitations upon reactor operation is deemed necessary to obviate j

! i the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety. (

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The final Commission Policy Statement established four criteria to define the scope of '

equipment and parameters to be included in the improved Standard Technical Specifications. These criteria were developed for licenses authorizing operation (i.e.,

operating reactors) and focused on instrumentation to detect degradation of the reactor coolant system pressure boundary and on equipment or process variables that affect the '

integrity of fission product barriers during design bases accidents or transients. The fourth criterion refers to the use of operating experience and probabilistic risk assessment to identify and include in the Technical Specifications structures, systems, and components shown to be significant to public health and safety. Nevertheless, these criteria, codified by 10 CFR 50.36, are the source of the technical specification requirements for safe storage of spent fuel. The staff gave consideration to these criteria as they apply to a plant with a reactor that is permanently shut down and defueled. A general discussion of these

considerations is provided below.

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' - Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that technical specification limiting conditions for operation must be established for " installed instrumentation that is used to detect. and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since no fuelis present in the reactor coolant system at the Connecticut Yankee facility, this criterion is not applicable.

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k 4-Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that technical specification limiting conditions for operation must be established for a " process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the design basis accident and transient analyses, and which

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are monitored and controlled during power operation. While this criterion was developed for operating reactors, there are some design basis accidents which continue to apply to a plant authorized only to handle, store, and possess nuclear fuel. The scope of DBAs applicable to a plant with a reactor that is permanently shut down and defueled is markedly reduced from those postulated for an operating plant. There are no operational transients which continue to apply.

Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that technical specification limiting conditions for operation must be established for structures, systems, or components (SSC's) that are part of the primary success path and which function or actuate to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into technical specifications only those SSC's that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria. While there are no transients which continue to apply to Connecticut Yankee, there are some design basis accidents which continue to apply to a plant authorized only to handle, store, and possess nuclear fuel. The scope of DBAs applicable to a plant with a reactor which is permanently shut down and defueled is markedly reduced from those postulated for an operating plant and there are no transients which continue to apply. The scope of DBAs applicable to the facility is discussed in more detail below.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that technical specification limiting conditions for operation must be established for structures, systems, or components (SSC's) which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of technical specification limiting conditions for operation. All of the accident sequences that previously dominated risk at Connecticut Yankee, due to the operation of the reactor, are no longer applicable with the reactor in the permanently shutdown and defueled condition.

Section 15 of the UFSAR described the DBA scenarios that were applicable to Connecticut Yankee Plant during power operations. However, as a result of the certifications submitted by Connecticut Yankee in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor in

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i accordance with 10 CFR 50.82(a)(2), the accident scenarios pertaining to plant operation 1 i

are no longer possible. Connecticut Yankee has revised the UFSAR (Change 30 dated January 1998) (and retitled it as the Decommissioning Updated Final Safety Analysis Reportl (DUFSAR)) to identify which accidents no longer apply.

Two accidents identified in the FSAR are applicable at Connecticut Yankee, as described in UFSAR Sections 15.5.1 and 15.5.2 and in the proposed technical specifications change. In the May 30,1997 submittal, the licensee summarized the results of analyses and a safety 1 assessment for the proposed changes, including the results of the following calculations:

SFP-97-1575-DY, Revision 1, " Decay Heat and Heatup Rate Analysis for the Connecticut Yankee SFP"; XX-XXX-60RA, Revision 1, " Radiological Assessment of a Spent Fuel Shipping Cask Drop in the CY Spent Fuel Pool" (which provides the radiological assessment of a fuel handling accident); and, CYRESIN-01578-RY, Revision 0, " Radiological )

Consequences from a Resin Fire". Although Connecticut Yankee is presently not licensed i

to permit lifting a shipping cask over the spent fuel pool, the licensee provided this evaluation to compare the consequences with those of a fuel handling accident. '

Section 15.5.1 provides an analysis of a postulated release from a subsystem or component  !

(waste liquid, waste gas, solid waste system and other decommissioning activity). The licensee considered failures in the waste liquid system and determined that failure of the aerated liquid waste disposal system does not present undue hazard to the public. During reactor operations, the waste gas decay tanks (WGDT) were used to store radioactive gases and permit their decay to reduce or prevent the normal release of radioactive materials to the atmosphere. The radioactive contents of the WGDTs were principally the noble gases krypton and xenon, the particulate daughters of some of the krypton and xenon isotopes, and trace quantities of halogens. These noble gases were generated from fission during operation of the reactor. Since the reactor is permanently shut down, such gases are no longer generated at the Connecticut Yankee facility. A gaseous release due to a failure of the waste gas system may contain trace quantities of dissolved noble gases. With the significant decay of noble gases and iodines since reactor shutdown in July 1996, the only volatile nuclide of any dose significance that could be released is Kr-85. The licensee completed an analysis assuming the release of the remaining noble gases on site as of July 1997. The calculated doses beyond the exclusion area boundary were significantly lower than the reference value in 10 CFR Part 100, and were significantly lower than those in the EPA Protective Action Guides.

Similarly, the licensee analyzed the release of airborne radioactivity due to an accident or i mishandling of solid radwaste materials. These types of accidents were considered in the I past but were not included in the UFSAR since the consequences were always bounded by the more significant doses from the gaseous waste system accident. However, with the small amount of radioactivity in the gaseous waste system for the permanently defueled condition, the licensee stated that the resin container accident is the bounding airborne radioactivity accident at Connecticut Yankee for decommissioning activities. The analysis of the resin accident found that the bounding airborne release occurred from an accident i

associated with a container of radioactive resins from the decontamination of the reactor '

coolant system and connected systems. The amount of curies assumed to be contained in  ;

the resin container was based on the maximum curies that the Department of Transportation shipping limits and NRC Class C buriallimits allow. One percent of this

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activity was assumed to be released to the environment, which agrees with the NRC l

guidance in NUREG 0782, Environmental Impact Statement on 10 CFR 61, " Licensing Requirements for Land Disposal of Radioactive Waste." No credit was taken in the analysis for filtration, confinement inside a building, or plate out of particulates on building surfaces.

The dose was calculated using two methods: one method used the NRC approved dose conversion factors from Regulatory Guide 1.109; and, the second method used dose f actors in accordance with EPA 400, " Manual of Protective Action Guides and Protective Actions for Nuclear incidents." The calculated doses beyond the exclusion area boundary were a small fraction of the reference value in 10 CFR Part 100, and were lower than the reference values established by the EPA Protective Action Guides. The potential airborne activity released from accidents that may occur during other decommissioning activities was bounded by the resin container accident.

Section 15.5.2 discusses the design basis fuel handling incident, and addresses two cases:

the drop of a single assembly inside the containment, and in the spent fuel pool. Fuel will no longer be handled in the containment at the Connecticut Yankee facility. The possibility of a fuel handling accident in the spent fuel poolis low due to the administrative controls and physicallimitations imposed on fuel handling operations. Nonetheless, a postulated fuel handling accident in the fuel building remains applicable.

The accident analysis provided in UFSAR Section 15.5.2 evaluated the consequences of a postulated drop of one fuel assembly onto the fuel racks. The number of ruptured fuel rods that would result depends on several variables including the kinetic energy at impact and fuel assembly orientation during impact. However, to assure that the limiting case is considered, the licensee assumed that all rods in the dropped assembly fail upon impact, resulting in the release of the gap activity. The assemblies were decayed for 221 days from the shutdown on July 22,1996. The isotopes that were assumed to remain were 1-129 and Kr-85, While the decontamination factor for any remaining iodine in the spent fuel water would be significant (DF= 100), the licensee calculated doses assuming both filtration and no filtration (DF= 1). No credit was taken for the spent fuel building ventilation system, resulting in an assumed instantaneous, unfiltered ground level release from the fuel building.

As described in License Amendment 188, the analysis of the fuel handling accident conducted by the NRC considered the plant specific mitigative features and assumed a bounding value of 60,000 MWD /T for Nel burnup versus the actual maximum value of 49,956 MWD /T used in the licerm . .nalysis. The NRC's analysis resulted in dose consequences limited to 10% o." W 40 CFR 100 limits. Because more than 500 days has elapsed since the last reactor opewtion, the consequences of a postulated fuel handling accident at the Connecticut Yankee facility will be even lower than those calculated in the previous analysis, in the May 30,1997 submittal, the licensee evaluated the postulated drop of a fuel handling cask onto fuelin the spent fuel pool. In the cask drop analysis, it is assumed that 157 assemblies (equivalent t'o one full core) are damaged. The licensee provided this analysis for comparison even though the present license does not allow the movement of a cask l

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over the spent fuel. As in the fuel handling analysis, the licensee assumed 221 days for l

fuel assembly decay and that the dominant isotopes that remained for a gaseous release were 1-129 and Kr-85. The analysis evaluated the thyroid dose from l-129 and the skin I

dose from Kr-85. The resulting offsite doses calculated for the cask drop accident are greater that the doses for the fuel handling accident, but remain well below the values specified in 10 CFR 100.

In addition to the fuel handling accident, the licensee conducted an analysis to calculate the reactivity effects of a mispositioned fuel assembly. This analysis was performed to support Amendment 188 to the Connecticut Yankee Technical Specifications. The analysis ,

considered inadvertent misloading of an assembly with a burnup and enrichment I i

combination outside the acceptable areas and a temperature increase above 150*F. Based

! upon the most reactive condition, a boron concentration of 300 ppm was determined to be adequate to mainta.in a 5%Ak/k safety margin to criticality.

l The licensee also considered the consequences of a loss of forced cooling to the spent fuel pool. The heat load in the spent fuel poolis much less than the design value, due to the decay time that has elapsed since reactor shut down (more than 500 days as of May 30, ,

1998). The rate of water loss from boiling and evaporation is low. Assuming the maximum i l allowable spent fuel pool temperature (150*F) and the minimum allowable SFP water level l (20 feet above the top of the fuel),it would take more than 13 days from the time a total loss of forced SFP cooling and makeup functions occurred to the time the SFP water level decreased to 10 feet above the fuel. A 10 foot water shield above the fuelis sufficient to  ;

protect personnel from excessive radiation exposure and provides an adequate heat sink for I the fuel. The 13 day time period is sufficient to provide an alternate source of forced cooling or to establish makeup flow to the SFP.

In summary, most of the accident scenarios described in UFSAR Section 15 are no longer possible due to the termination of reactor operations and the removal of fuel from the reactor at the Connecticut Yankee facility. The remaining postulated design basis accidents I used in the development of the proposed DTS are a fuel handling accident in the spent fuel pool, an extended loss of forced cooling to the spent fuel pool, and radioactive waste t

system failures.

The following is a section-by-section discussion of the proposed revision to the Connecticut Yankee License and Appendix A Technical Specifications. The majority of the changes reflect the limitations and requirements appropriate to the permanent defueled condition of the plant. In addition, because this proposed amendment affects most pages in the Operating License and the Technical Specifications, some changes were made to correct typographical errors or grammar; or to provide consistent terminology, consistent format, clarity, or brevity; or to update the index (Table of Contents) to reflect the proposed changes. Each section of the license and current TS (CTS) is listed and the proposed

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FACILITY OPERATING LICENSE License Condition 2.B.(3):

Change " reactor startup" to " reactor startup (possession only),"

and change " reactor instrumentation" to " reactor instrumentation (possession only)". The HNP may have such sources in its possession from previous operational cycles. The cha reflects the permanent defueled plant condition by only allowing possession, not use of these sources. The staff finds the changes on the use of sources to be acceptable.

License Condition 2.C.(1), Maximum Power Level: Change "The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1825 megawatts (thermal)." to "The licensee is not authorized to operate the reactor. Fuel may not be placed in the reactor vessel." Pursuant to 10 CFR 50.82(a)(1)(l) and 10 CFR 50.82(a)(1)(ii), CYAPCo has certified that the plant has permanently ceased operations and the reactor _ has been permanently defueled. The change maintains this present plant configuration by prohibiting the placement of fuelin the reactor vescc!. The staff finds tus changes to restrict reactor operation and placement of fuelin the vessel to be acceptable.

License Condition 2.C.(2), Technical Snecifications: Change " Amendment No.190," to

" Amendment No.193,". (the blank to be filled in prior to issuance).

The staff finds this editorial change reflecting the new amendment to be a ceptable.

License Condition 2.C.(3), Deleted: Change "I(3) Deleted per Amdt. 29,10-24-78.]" to

"(3) to be Deleted by acceptable. Amendment No. 29." The staff finds this editorial change for consistency License Condition 2.C.(4), Fire Protection: Replace the second paragraph with: "The licensee may make changes to the fire protection program without NRC approvalif these changes do not reduce the effectiveness of fire protection for facilities, systems, and equipment which could result in a radiological hazard, taking into account the decommissioning plant conditions and activities." Because the plant has permanently ceased operations and the reactor has been permanently defueled, there no longer is the need to achieve and maintain safe shutdown of the reactor in the event of a fire. The change rehects that, in the event of a fire, the Fire Protection Program is now focused on potential radiological hazards, taking into account the decommissioning plant conditions and activities. The staff finds this change to be acceptable.

License Condition 2.C.(5), Physical Protection: No change.

License Condition 2.C.(6) , Intearated imolementation Schedule: No change.

License Condition 2.C.(7), Fuel Movement: The following, item C.(7), is added: "The movement of special nuclear material used as reactor fuel into the containment is prohibited." The plant has permanently ceased operations and the reactor has been permanently defueled. The new license condition maintains this present plant configuration by prohibiting the movement of fuelinto the containment.

acceptsble. The staff finds the change to be APPENDIX A TECHNICAL SPECIFICATIONS

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CTS Ganarmi and inder I

t This license amendment retyped the entire Operating License and Technical Specifications Sections 1 through 5 and the Bases to assure that pages that were supposed to be retained i

will not be inadvertently discarded. Since this will be a new amendment, the revision bars l

were remcved from Sections 1 through 5 and the Bases. The staff finds this change to be acceptable. > (

d The licensee ma' e editorial changes to the Index to reflect the changes made in the body of ,

the CTS. Sections deleted were indicated by " DELETED" and, if all subsections of a i

technical specification were deleted, then only the major section was shown to be deleted and the subsections were removed from the index. Similarly, if a table or figure was  ;

deleted, the table or figure was removed from the index. The licensee retained the numbering s,equence to preclude the need to renumber the cross-references. The staff finds the proposed changes to the index to be acceptable. {

L CTS Section 1.0. Definitions i

This section of the current Connecticut Yankee TS contains the definitions for terminology that is unique to plant operations. Many of the definitions of the CTS applied to reactor i operations and are no longer applicable to the Connecticut Yankee facility. Similarly, ,

references to operational modes 1 through 5, and the frequency notations of "startup" and

" prior to each reactor criticality" are no longer applicable to a plant that has been permanently defueled. The definitions deleted were not referenced by the remaining ,

technical specifications. The staff finds the proposed deletion of certain definitions to be appropriate for the permanently shutdown and defueled status of the plant. The staff finds the' proposed changes to the Definitions to be acceptable. .

1 CTS Section 2.0, Safety Limits and Safety System Settings Section 2 of the CTS contains " safety limits" and "li,miting safety system settings." in i accordance with 10 CFR 50.36(c)(l), safety limits sie limiting parameters necessary to protect the physical barriers that guard against the uncontrolled release of radioactivity from

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a nuclear reactor, if a parameter exceeds the specified safety limit, the reactor must be l-  !

shut down and operation ma'y not resume until authorized by the NRC. Limiting safety l system settings are values of various parameters associated with the nuclear steam supply  !

system (NSSS) at which automatic protective action is needed during normal operations or anticipated transients to prevent violation of the safety limits. ,

1 The CTS contain two safety limits. CTS 2.1.1 concerns the reactor core and sets limitations on the reactor thermal power, pressurizer pressure, and loop inlet temperature.

l These limits prevent damage to the fuel cladding during reactor operation that could result i

in the release of fission products to the reactor coolant system. CTS 2.1.2 places a

! limitation on the pressure in the reactor coolant system. This limitation prevents potential damage to the reactor coolant system pressure boundary that could result in the release of l fission products to the containment atmosphere. The limiting safety system settings are contained in CTS 2.2.1. This specification establishes limits on the set points of the reactor protection system (RPS). The RPS monitors various parameters associated with reactor

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operation and initiates a shutdown of the reactor if the settings are exceeded during normal operation or anticipated operational occurrences.

The Connecticut Yankee f acility is permanently shut down and fuel has been removed from the reactor vessel and placed in the spent fuel storage pool. The facility operating license no longer authorizes operation of the reactor or placement or retention of fuelin the reactor. '

Since the reactor is not in operation at the Connecticut Yankee facility and the regulations prohibit such operation in the future, the licensee states that CTS 2.1.1 and 2.1.2 are no longer applicable and proposes to delete them from the DTS. Also, because the reactor at the Connecticut Yankee facility is no longer authorized to operate, the RPS no longer serves a useful function. Therefore, CYAPCo also proposes to delete CTS 2.2 in the DTS. The staff finds the proposed deletion of the safety limits to be appropriate due to the permanently shutdown and defueled status of the reactor plant and the dcsciivation of the reactor coolant system and nuclear steam supply system. Since the reactor core and reactor coolant system p!sy no role in the safe storage of irradiated fuel at Connecticut Yankee, the staff finds the proposed changes to CTS Section 2.0 to be acceptable.

CTS Section 3.0 and 4.0, Limiting Conditions for Operation and Surveillance Requirements in accordance with 10 CFR 50.36(c)(2), limiting conditions for operation (LCOs) specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The LCOs typically place restrictions on the availability of safety equipment needed to prevent or mitigate a postulated DBA or on process variables necessary to preserve the initial conditions assumed in analyses of postulated design basis events. As 1 discussed above,10 CFR 50.36(c)(2)(ii) defines four criteria for establishing limiting {

conditions for operations. Associated surveillance requirements help to ensure that the j specified equipment and parameters are maintained within the limits specified in the LCOs. '

As discussed previously, only a limited set of postulated design basis accidents remain applicable to the Connecticut Yankee facility with its reactor in the permanently defueled state. As a result, CYAPCo determined that most of the LCOs and accompanying surveillance requirements contained in the CTS were inappropriate for retention in the DTS.

The subsections of CTS Sections 3.0 and 4.0 are discussed below.

CTS Subsection 3/4.0, Applicability This subsection contains specifications that have generic applicability to the LCOs and surveillance requirements. Due to the limited number of LCOs remaining in the DTS, a number of the CTS provisions in this section are no longer necessary for or applicable to the Connecticut Yankee facility. In subsections 3.0.1, 3.0.4, 4.0.1, 4.0.4, and in the associated bases for thase specifications, the licensee changed references to ,

" OPERATIONAL MODES" and " shutdown is required to place the facility in a MODE or

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condition". Deleting references to OPERATING MODES, refueling, plant startup and plant shutdown is consistent with the plant having permanently ceased operations and the {

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1 reactor being permanently defueled. Substituting the words "specified applicable condition" for " OPERATIONAL MODES or other conditions specified" retains the intent of the applicability for the conditions specified in each Limiting Condition for Operation. The staff finds the changes to be acceptable.

Subsection 3.0.3 was deleted in its entirety, along with the associated bases. The deleted specification has requirements specific to operating modes and requirements associated with fuelin the reactor vessel and/or power operation of the facility. Because the plant has permanently ceased operations and the reactor has been permanently defueled, this specification is no longer required. The staff finds the changes to be acceptable.

STS 3.0.4 prohibits entry into an operating mode or other specified condition in the applicability of an LCO unless the surveillances have been met within their specified frequency. This prohibition is subject to certain conditions. Since the reactor at the Connecticut Yankee facility is permanently shut down and defueled, there will no longer be any changes in operating condition or mode. Therefore, the licensee has proposed not to include this specification in the DTS. The staff has determined that, due to the permanently shutdown and defueled status of the Connecticut Yankee facility, this subsection of the STS is not necessary for safe operation or maintenance of the plant, and finds the proposal not to include it in the Connecticut Yankee DTS to be acceptable.

The change to 3.0.2 is editorial. The movement of Specifications 4.0.1 and 4.0.2 to Page 3/4 0-2 is editorial and is being done for clarifi<:ation, since these two sections should be on a page labeled " Surveillance Requirements." The change to B4.0.1/B4.0.5 reflects the deletion of B4.0.5. In Subsection B3.0.4, the licensee proposed the following changes:

(i) in the first paragraph, change "on MODE changes when" to "on changes from a specified applicable condition when"; (ii) delete the second sentence; (iii) third sentence, change

" ensure that facility operation is not initiated or that higher MODES of operation are not entered when" to " ensure that the facility is not changed from a specified applicable condition when"; (iv) fourth sentence, delete "before or after a MODE change"; (v) fifth sentence, change "into an OPERATIONAL MODE or other specified condition may" to into a specified applicable condition may"; and (vi) sixth sentence, delete "before plant startup" The licensee also proposed to delete the second paragraph. The staff finds these changes to be acceptable.

In accordance with 10 CFR 50.36(c)(3), surveillance requirements are related to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operations will be met. Since there are relatively few remaining LCOs, the licensee has proposed to greatly reduce the number of surveillance requirements.

In Section B4.0.1/B4.0.5, the licensee proposed the following change in the first sentence:

change "Soecifications 4.0.1 throuah 4.0.5" to "Soecifications 4.0.1 throuah 4.0.4". In Section B4.0.1, the licensee proposed the following changes: (i) in the first sentence, change "the OPERATIONAL MODES or other conditions" to "the specified applicable conditions"; (ii) in the second sentence, change "a MODE or other specified condition" to "a specified applicable condition"; and, (iii) delete the third and fourth sentences. In Section B4.0.2, the licensee proposed to delete the third and fourth sentences. In Section B4.0.3, the licensee proposed in the second paragraph, first sentence, to change "less than 24

hours or a shutdown is required to comply with ACTION requirements, e.g., Specification 3.0.3, a 24-hour allowance" to "less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, a 24-hour allowance". In the second paragraph, third sentence, delete "a shutdown is required to comply with ACTION requirements or before other". In Section B4.0.4, the licensee proposed the following changes: (i) in the first paragraph, first sentence, change "an OPERATIONAL MODE or other specified condition" to "a specified applicable condition"; (ii) second sentence, change "into a MODE or condition for" to "into a specified applicable condition for"(iii) delete third sentence; and, (iii) delete the second and third paragraphs. These changes reflect the fact that operational modes are no longer applicable at the HNP. The staff finds these changes to be acceptable.

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The licensee proposed deleting Subsections 4.0.5 and 4.0.6, including the associated  !

tables, figures and bases, and the References (Page 83/4 0-7). These subsections are  !

associated with fuelin the reactor vessel and/or power operation of the facility. Because  !

the plant has permanently ceased operations and the reactor has been permanently i defueled, these subsections are no longer required. Specifically, Section 4.0.5 is concerned I with the ASME Section XI inspection, examination and testing requirements for Class 1, 2 and 3 components. Because the plant has permanently ceased operations and the reactor has been permanently defueled, the remaining functional systems covered by this Specification are the service water system (SWS) and the spent fuel cooling system (SFCS). j The SWS design pressure is 110 psig and the design temperature is 120*F. The SFCS design pressure is 200 psig and the design temperature is 150*F. These systems, while not designed to ASME Ill, are classified as ASME Section ill, Code Class 3 (

Reference:

UFSAR Table 3.2-1).

Inspections, examinations and tests verify the integrity of high pressure and temperature  ;

systems. The remaining functinnal systems at the HNP are relatively low pressure and temperature systems. Therefore, high temperature / pressure erosion and corrosion is not expected. In ASME Section XI, Table IWD-2500-1 (" Test and Examination Categories"),

1983 Edition with Sumner 83 Addenda, the SWS examination requirements are delineated  !

in items D2.10 through D2.60 and the SFCS examination requirements are delineated in items D3.10 through D3.60. In both cases the examinations consist of a system pressure test with a visualinspection for leakage (ltem D2.10 and D3.10) and a visualinspection of supports (items D2.20 through D2.60 and items D3.20 through D3.60). There are no requirements for surface or volumetric examinations. Thus, while ASME Section XI requires an additional record keeping woridoad, from a technical standpoint the code requirements are being met daily by Operations and other plant personnel. The licensee stated that modified ISI and IST requirements customized to the defueled plant condition will be incorporated in the Technical Requirements Manual to assure component and system readiness in performing their required functions. Furthermore,10 CFR 50.55a(g), which requires an examination and testing program in accordance with ASME Section XI,is predicated on power operations. Thus, following the submittal of the 10 CFR 50.82 certification, ASME Section XIis no longer applicable to the HNP. Therefore, based on the above, TS 4.0.5 is no longer needed. The staff finds this change to be acceptable.

Subsection 4.0.6 is concerned with maintaining integrity of auxiliary feedwater system piping that is considered high energy piping (i.e., operating pressure 2 275 psig and/or operating temperature 2 200*F). Similarly, the references (Page B3/4 0-7) which deal with

.. -- .. -. - -.- -- . ~-. - -.- - - . - - - . - - - - . . - . ._ -

i EQ and HELB are no longer applicable and are no longer required. Because the plant has permanently ceased operations and the reactor has been permanently defueled, this auxiliary feedwater system piping will no longer see these operating pressures and/or -

temperatures. Therefore, this specification and the references are no longer required. The {

staff finds the change to be acceptable.

CTS Subsection 3.1 and 4.1 Reactivity Control Systems The purpose of Subsection 3/4.1 was to assure sufficient shutdown margin exists so that the reactor can be made subcritical from all operating conditions, the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. CTS 3.1 and 4.1 contains LCO requirements which must be satisfied l

to limit the reactivity condition of the reactor core. LCOs in this section include limitations '

on shutdown margin, moderator temperature coefficient, minimum temperature for criticality, boration systems and flow paths, and movable control element assemblies.

Because the plant has permanently ceased operations and the reactor has been permanently i defueled, the reactor related accidents and reactivity conditions are no longer possible.

Therefore, these specifications are no longer applicable and this subsection is no longer required. CYAPCo proposed to delete the entire subsection, including the associated tables and bases.

This specification is no longer necessary for safe operation or maintenance of the plant, and the staff finds its deletion to be acceptable.

CTS Subsection 3.2 and 4.2, Power Distribution Limits The purpose of Subsection 3/4.2 was to ensure power distribution in the reactor remained within specified limits to preserve fuelintegrity during operation. The specifications provide assurance of fuel integrity during normal operation and postulated design basis transient and accident conditions. LCOs in this section include limitations on core power axial offset, linear heat generation rate, power peaking factors, quadrant power tilt ratio, and core thermal-hydraulic conditions. These limitations ensure that the integrity of the fuel cladding is maintained during normal reactor operations and ' anticipated transients and that the initial i conditions assumed in the analyses of postulated accidents affecting the reactor core remain valid. Because the plant has permanently ceased operations and the reactor has been permanently defueled, the reactor related accidents and core power conditions are no

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longer possible. Therefore, these specifications are no longer applicable and this subsection is no longer required. CYAPCo proposed to delete the entire subsection, including the associated table and bases.

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This specification is no longer necessary for safe operation or maintenance of the plant; the staff finds its deletion to'be acceptable.

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l CTS Subsection .3.3 and 4.3, instrumentation This subsection of the CTS contains LCO's related to a wide variety of instrumentation systems 3/4.3.1 and 3/4.3.2 cover the instrumentation associated with the reactor trip system and the engineered safety features actuation system. These systems are designed l to shut down the reactor or initiate automatic protective actions when parameters exceed selected limits. The systems function to prevent or mitigate the consequences of postulated accidents that could result in damage to the reactor fuel cladding or the reactor coolant pressure boundary. Reactor operations have been terminated at the Connecticut Yankee facility and the reactor has been permanently defueled. Therefore, the postulated accident scenarios requiring actuation of these systems are no longer possible. The staff finds the deletion of these subsections to be acceptable.

Similarly, the following subsections specify requirements for instruments associated with fuel in the reactor vessel and/or power operation of the facility. Specifically, the purpose of Subsection 3/4.3.3.1 was to ensure reactor coolant system (RCS) leakage detection inside containment and to assess the integrity of the reactor coolant pressure boundary during power operations. The purpose of Subsection 3/4.3.3.2 was to ensure accurate neutron flux readings were obtained. The purpose of Subsection 3/4.3.3.5 was to monitor accidents affecting the reactor coolant system (RCS) and/or containment. Because the plant has permanently ceased operations and the reactor has been permanently defueled.

the accidents that these instruments would monitor are no longer possible. Therefore, this instrumentation is no longer needed. The staff finds the changes to be acceptable Subsection 3/4.3.3.6 had been previously deleted in Amendment No.179.

The purpose of Subsection 3/4>3.3.9 was to assure that boron levels inside the reactor core precluded criticality. The purpose of Subsection 3/4.3.4 was to ensure protection against flooding of safety-related equipment that is associated with power operation of the reactor coolant system (RCS) or RCS/ containment accident mitigation. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility, these subsections are no longer necessary for safe operation or maintenance of the plant, and their deletion is acceptable.

The licensee proposed to retain severalinstrumentation subsections and tables, subject to certain changes. In Subsections 3/4.3.3.3, Seismic instrumentation, and 3/4.3.3.4, Meteorological Instrumentation, the licensee proposed to delete ACTION "b," which refers to deleted Specification 3.0.3. In 3/4.3.3.7, Radioactive Liquid Effluent Monitoring Instrumentation, the licensee proposed to delete ACTION "c," which refers to deleted Specification 3.0.3. In the footnote, " outages are permitted" was changed to " outages of monitoring channels are permitted". The change to the footnote clarifies that " outages" i

refers to the monitoring channels, not plant outages. The staff finds these changes to be acceptable. '

In TABLE 3.3-9, Radioactive Liquid Effluent Monitoring Instrumentation, the licensee proposed to delete items 1.b and 3.b, ACTION 47, and associated notes. These changes refer to steam generator blowdown. In TABLE 4.3-7, Radioactive Liquid Effluent Monitoring instrumentation Surveillance Requirements the licensee proposed to change "B."

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to "b."in item 3.B. The change is an editorial change for consistency. The licensee proposed to delete items 1.b and 3.b, NOTATION (5), and the associated notes, and to delete "and steam generator blowdown line"in the note associated with NOTATION (311.

This set of changes refers to steam generator blowdown. Because the plant has permanently ceased operations and the reactor has been permanently defueled, it is no longer necessary to monitor steam generator blowdown. The staff finds these changes to be acceptable. '

in Subsection 3/4.3.3.8, Radioactive Gaseous Effluent Monitoring instrumentation, the licensee proposed to change " Alarm / Trip" to " Alarm"in the lead paragraph and in ACTION "a." The licensee proposed to delete ACTION "c," which refers to deleted Specification 3.0.3. In the footnote, the licensee proposed to change " outages are permitted" to

" outages of monitoring channels are permitted". The change to the footnote clarifies that

" outages" refers to the monitoring channels, not plant outages. With respect to the change of " Alarm / Trip" to " Alarm", the automatic isolation of waste gas system releases is no longer necessary. The purpose of the waste gas system is to collect and store for decay the radioactive gasses generated in the RCS during reactor operation. Because the plant has permanently ceased operations and the reactor has been permanently defueled, this system is no longer operated and is no longer required. The staff finds these changes to be acceptable.

In TABLE 3.3-10, Radioactive Gaseous Effluent Monitoring instrumentation, the licensee proposed to delete "and Automatic Termination of Waste Gas System Releases"in item 1.a. With respect to item 1.a, the deletion refers to the automatic termination of waste gas system releases. The licensee proposed to delete item 1.b, which refers to the lodine Sampler. In TABLE 4.3-8, Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Requirements, the licensee proposed to delete item 1.b. In NOTATION (3)a.

the licensee proposed to change " Alarm / Trip" to " Alarm". The licensee proposed to delete the note associated with NOTATION (3)a, which refers to waste gas releases.

In considering the changes to Table 3.3-10 and 4.3-8, the licensee noted that the plant was shut down on July 22,1996. Except for 1-125 (half-life ~ 59.5 days),1-129 (half-life

- 1.6E7 years), and Kr-85 (half-life - 10.3 years), the spent fuelinventory of the dose-contributing radioactive iodine and noble gas isotopes have decayed more than 20 half-lives since shutdown (i.e., less than 0.0001% of the original amount remains). In addition, the definition for " Dose Equivalent 1-131" (" Standard Technical Specifications. Westinghouse Plants," NUREG-1431) does not include I-125 and 1-129 in the dose assessment due to their negligible inventory in the spent fuel. Due to the decay time since reactor shutdown, iodine sampling of the main stack is no longer necessary. Except for Kr-85, the noble gas nuclides that contribute to a whole body dose have also decayed to a negligible amount. CYAPCO has performed fuel handling and cask drop accident dose calculations which conclude that doses (i.e., whole body and thyroid) at the Exclusion Area Boundary are a small fraction of the 10 CFR 100 dose limits and the EPA PAGs. With respect to the change to NOTATION (3)a and the note associated with NOTATION (3)a, the automatic isolation of waste gas system releases is no longer necessary. Because the plant has permanently ceased operations and the reactor has been permanently defueled, this system is no longer operated and is no longer required. The staff finds the changes to Tables 3.3-10 and 4.3-8 to be acceptable, l

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j CTS Subsection 3.4 and 4.4, Reactor Coolant System The purpose of Subsection 3/4.4 was to ensure that the fuelin the reactor vessel and/or power operation of the facility was maintained and/or operated in a safe manner. The j

specifications provide the requirements for reactor coolant system components to assure 1) adequate reactor core heat transfer capability under all shutdown, operating, transient, and emergency conditions, 2) uniform RCS boron concentration during boration or dilution I

evolutions, 3) over pressure protection (heatup, cooldown, and pressure-temperature limitations),4) protection of the functionalintegrity of the reactor materials, and 5) the integrity of the reactor vessel and reactor coolant pressure boundary. Since the reactor at the Connecticut Yankee facility has been permanently shut down and defueled, the functions of the reactor coolant system components are no longer required to prevent or mitigate the consequences of postulated accidents. Therefore, the CYAPCo proposes to delete the entire subsection, including associated tables, figures and bases.

Because the plant has permanently ceased operations and the reactor has been permanently defueled, the operating conditions of the reactor core and reactor coolant system are no longer of concern. Further, this subsection specifies temperature and pressure related conditions during RCS heatup and cooldown to assure protection of the reactor coolant pressure boundary, which are based upon assuring compliance with 10 CFR 50 Appendix G, as invoked by 10 CFR 50.60. On July 29,1996, the NRC issued the final decommissioning rule and amended 10 CFR 50.60 to exempt facilities which have submitted 10 CFR 50.82(a)(1) certifications from 10 CFR 50 Appendices G and H. Therefore, these LCOs are no longer applicable. The staff finds deletion of this specification to be acceptable.

CTS Subsection 3.5 and 4.5, Emergency Core Cooling These LCOs are concerned with the operation of various emergency core cooling systems.

These systems include the centrifugal charging pumps, high pressure safety injection pumps, low pressure safety injection pumps, residual heat removal pumps, and residual heat I

! removal heat exchangers, associated valves, and the refueling water storage tank. The i purpose of Subsection 3/4.5 was to ensure operability of the emergency core cooling system (i!CCS). The limitations on the operation of this equipment ensure that cooling can l be provided to the reactor following a postulated loss of coolant accident. Because the J plant has permanently ceased operations and the reactor has been permanently defueled.

l the accidents the ECCS would mitigate are no longer possible. Thus this subsection is no i longer required. Therefore these specifications are no longer applicable and CYAPCo proposed to delete the entire subsection, including associated tables, figures and bases.

The staff agrees that this subsection is no longer necessary for safe operation or maintenance of the plant and finds its deletion to be acceptable.

! CTS Subsection 3.6 and'4.6, Containment The purpose of Subsection 3/4.6 was to ensure containment of radioactive releases due to design basis accidents with fuelin the reactor core or during in-containment movement of spent fuel. The LCO requirements serve to ensure the integrity of the primary containment.

The primary containment serves to limit the release of radioactive material to the l

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environment in the event of postulated accidents that release radioactive materials from the reactor coolant system. Because the plant has permanently ceased operations and the reactor has been permanently defueled, the in-containment design basis accidents are no longer possible. Therefore, these specifications are no longer applicable and CYAPCo proposed to delete the entire subsection, including associated bases.

Since the reactor at the Connecticut Yankee facility has been permanently shut down and defueled, there are no remaining postulated accidents which require the integrity of the prin.ary containment to be maintained. The staff has determined that this subsection is no longer necessary for safe operation or maintenance of the plant, and finds its deletion to be acceptable.

CTS Subsection 3.7 and 4.7, Plant Systems This subsectio' n of the CTS contains LCOs related to a variety of plant systems that were important to plant operations. These systems are designed to support normal power operations and the shutdown to cold conditions, or mitigate the consequences of postulated anticipated transients and accidents that could result in damage to the reactor fuel cladding j or the reactor coolant pressure boundary. Plant operations have been terminated at the Connecticut Yankee facility and the reactor has been permanently defueled. Therefore, the postulated scenarios requiring operation of these systems are no longer possible. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility, these subsections are no longer necessary for safe operation or maintenance of t,he plant, and can be either modified to reflect the shutdown status, or deleted entirely.

In Subsection 3/4.7.5, Sealed Source Contamination, the licensee proposed to delete ACTION. "b," which refers to deleted Specification 3.0.3. In Surveillance 4.7.5.2.c, the licensee proposed to delete "within 31 days prior to being subjected to core flux or installed in the core". The deleted portion of Surveillance 4.7.5.2.c is no longer needed because it related specifically to power operations and fuelin the reactor. The staff finds these changes to be acceptable.

The following CTS sections were deleted, including associated tables and bases. These subsections are associated with fuelin the reactor vessel and/or power operation of the facility. Because the plant has permanently ceased operations and the reactor has been permanently defueled, these subsections are no longer required. Specifically, the purpose of Subsection 3/4.7.1, Turbine Cycle, was to ensure that the effects of in-containment accidents (i.e., LOCA, MSLB, MFWLB) did not exceed containment design limits or off-site dose limits. Because the plant has permanently ceased operations and the reactor has been permanently defueled, accidents affecting the steam system are no longer possible and the specification is no longer needed. The purpose of Subsection 3/4.7.2, Steam Generator Pressure / Temperature Limitation, was to ensure that pressure-induced stresses within the steam generator did not exceed established limits. Because the plant has permanently ceased operations and the reactor has been permanently defueled, there is no longer a need to ensure steam generator integrity. Therefore, the limits are no longer needed. The staff finds these changes to be acceptable.

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The purpose of Subsection 3/4.7.3, Service Water System (which was applicable only in  !

MODES 1 through 4), was to ensure that sufficient cooling water was provided to safety-related systems during normal operation and accident conditions. The service water system provided cooling water to the SFP cooling system. The Technical Specification amendment request contains Subsection 6.8.1.g, which requires that the licensee provide a spent fuel pool cooling and makeup monitoring program to provide reasonable assurance i

that these systems are capable of tolfilling their intended functions. In addition, the licensee has begun installation of en a!temate heat sink for the SFP cooling system, which does not rely on service water. By letter dated June 18,1998, the licensee committed to maintain forced SFP cooling from the service water system until the alternate cooling system has been satisfactorily tested. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility and the i

significant reduction in spent fuel decay heat load, Subsection 3/4.7.3 is no longer necessary for safe operation or maintenance of the plant, and its deletion is acceptabic.

The purpose of Subsection 3/4.7.4, Snubbers was to ensure that the structuralintegrity of applicable safety-related systems was maintained fo!!owing a seismic or other event initiating dynamic loads. Snubbers are located on the following safety-related systems:

I RCS, CVCS, RHR system and feedwater system (safety-related portion). None of these safety-related systems are required to be operable because the plant has permanently ceased operations and the reactor has been permanently defueled. The staff has i determined that deletion of this subsection is acceptable. ,

t l Subsections 3/4.7.6, 3/4.7.7, and 3/4.7.8 had been previously deleted in Amendment No.179.

I The purpose of Subsection 3/4.7.9, Feedwater Isolation Valves, was to ensure that the I effects of in-containment main feedwater line break accidents did not exceed containment design limits or off-site dose limits. The purpose of Subsection 3/4.7.10, External Flood Protection, was to ensure that certain actions were initiated to protect safety-related i equipment (i.e., safety-related equipment in the plant auxiliary building (PAB), the service water system and the emergency diesel generator (EDG) system) from flooding by the Connecticut River. The purpose of CTS Section 3/4.7.11, Primary Auxiliary Building Air i

' Cleanup System, was to ensure that the effects of in-containment accidents (i.e., LOCA, MSLB, MFWLB, fuel handling accident) did not exceed off-site dose limits. The purpose of CTS Section 3/4.7.12, Ultimate Heat Sink (which was applicable only in MODES 1 through 4), was to ensure that sufficiently cold water (i.e., s; 90*F) was provided to safety-related systems during normal plant operation and accident conditions. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility, these sections and subsections are no longer necessary for safe operation or maintenance of the spent fuel, and that their deletion is acceptable.

CTS Subsection 3.8 and'4.8, Electric Power Systerns The purpose of Subsection 3/4.8 was to ensure that sufficient power was available to supply the safety-related equipment. The subsection contains LCOs associated with the AC and DC power sources and distribution systems. The LCOs are intended to ensure that sufficient power is available to supply the safety related equipment required for 1) the safe i

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shutdown of the facility, and 2) the mitigation and control of accident conditions within the f acility. The licensee states that these LCOs are not applicable with the reactor in the perrnanently shutdown and defueled condition. CYAPCo states that the first item is no longer applicable because the plant has permanently ceased operations and the reactor has been permanently defueled. With respect to the second item, the defueled condition leaves only a loss of spent fuel pool forced cooling and the off-site dose consequences of a fuel handling accident. As discussed below, CYAPCo states that this specification is not needed to mitigate either of these postulated accidents.

The plant was shut down on July 22,1996 and as of May,1998 more than 500 days have passed since the shutdown. Thus, the heat load on the spent fuel pool cooling system is greatly reduced. Present cooling performance data as well as calculations d6monetrate that either the plate or the shell and tube heat exchanger has more than adequate heat removal capacity. In the event of a loss of forced cooling, calculations indicate that the spent fuel pool time to boilis greater than 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> based on an initial pool temperature of 150 F. A previous Safety Evaluation, performed for Ucense Amendment No.188, determined that cooling the SFP by allowing the SFP to boil and adding makeup water was acceptable in the event of a complete loss of the capability of using heat exchangers to remove heat from the SFP. In the event that boiling commences, the operators have in excess of 13 days to provide forced cooling and/or makeup before there is inadequate shielding provided by the water in the pool. Therefore, operability of spent fuel pool cooling does not require a power source to be immediately available. The licensee states that there is sufficient time to effect repairs to the cooling system or to establish rnakeup flow prior to SFP water level decreasing below 10 feet above the fuelin the event of a loss of cooling to the spent fuel pool. Since active safety systems are not contained in the proposed DTS, the licensee has proposed not to retain the specifications covering electrical power to support such spent fuel related systems in the DTS.

Since all fuel has been permanently removed from the containment, a fuel handling accident in containment is no longer possible. However, a fuel handling accident in the spent fuel building is still part of the HNP design and licensing basis. The licensee stated that, in the event of a loss of offsite power, procedures require the operators to stop any work in progress in the SFP and to store, in a safe condition, any items suspended from the crane until onsite emergency power is restored.

The tool used to handle spent fuel within the SFP contains a gripper that is totally mechanical and all operations are conducted under a minimum water depth of 7 feet. There is no electrical power required to maintain contact with the lifting device and item being handled within the pool. Therefore, by virtue of the design of the gripper and compliance with applicable procedures, a fuel handling accident inside the spent fuel building would not result from a loss of offsite power.

CYAPCo stated procedures were in place to establish onsite power in the event of a ' oss of Normal Power (LNP) and in the event of a loss of cooling to the Spent Fuel Pool. Foi a LNP, power can be made available within approximately one hour, if onsite power cannot be reestablished within approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a LNP, limited makeup water could be provided by gravity feed from a tank (available in approximately 30 minutes) or an unlimited supply of water could be provided via the diesel fire pump from the Connecticut River

C (available in approximately 30 minutes). Therefore, within approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the event start, the licensee would reestablish cooling and/or makeup to the spent fuel pool.

The licensee stated that the longest LNP that HNP has experienced has been less than 30 minutes.

The licensee stated that, although this subsection was no longer required, relevant portions will be retained in the Technical Requirements Manual (TRM) to provide an additional set of requirements for electrical systems that support spent fuel cooling that are formally controlled. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility and the extended length of time since final shutdown, sufficient time exists for the licensee to respond to this event before any challenge is presented to the integrity of a fission product barrier. The staff has determined that this subsection is not necessary for safe storage of fuel at the site and finds its deletion to be acceptable.

CTS Subsection 3.9 and 4.9, Refueling Operations This CTS subsection contains a number of specifications related to refueling operations. In a letter dated May 7,1998, the licensee supplemented the license amendment request submitted on May 30,1997 by requesting the addition of Technical Specification 3/4.9.16

(" Spent Fuel Pool Cooling - Defueled") to assure appropriate spent fuel pool temperature limits are maintained while spent fuelis stored in the spent fuel pool. No other changes were proposed for this subsection, and the LCOs and associated tablec, figures and bases were retained.

Technical Specification 3.9.15 limited the spent fuel pool temperature to 150'F during refueling operations. The basis for the 150 degree F limit was to limit the thermal stresses in the spent fuel pool concrete structure due to differential temperature across the internal and exterior surfaces of the walls and floors. However, TS 3/4.9.15 does not apply to the permanently defueled i,ondition of the plant. Proposed TS 3/4.9.16 reestablishes the 150 degree F temperature limit for the SFP. Since TS 3/4.9.16 preserves the limit previously established in Technical Specification 3.9.15, as approved by the NRC in License Amendment No.188, the staff finds the change to be acceptable.

(TS 3/4.9.15 also contained operability and surveillance requirements for the SFP cooling pumps and heat exchangers. The licensee added TS 6.8.1.g. as discussed below, to provide reasonable assurance that the SFP cooling and makeup systems will perform their intended functions.)

CTS Subsection 3.10 and 4.10, Special Test Exceptions This CTS subsection contains limiting conditions for operation to allow physics testing of the reactor and measurernent of control rod performance. Since the reactor at the Connecticut Yankee facility has been permanently shut dowr and defueled, this specification is no longer applicable, and the licensee proposes to delete it from the DTS. '

The staff agrees that this subsection is no longer necessary for safe operation or maintenance of the plant, and finds its deletion to be acceptable.

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CTS Section 3/4.11, Radioactive Effluents Specification 3/4.11 is provided to implement the requirements of Appendix 1 of 10 CFR Part 50, and to ensurc that the concentration of radioactive materials released in effluents from the site will be less than the concentration levels in 10 CFR Part 20, Appendix B. The licensee proposed to retain these specifications with changes that reflect the permanent shutdown condition of the plant. The following specifications are retained with editorial changes. In Sections 3.11.1.2, 3.11.2.2, 3.11.2.3, and 3.11.3, the licensee proposed to delete ACTION "b," which refers to deleted Specification 3.0.3. Further, in Section 3/4.11.2.2, the licensee proposed to change the title from " DOSE-NOBLE GASES" to

" DOSE, NOBLE GASES". The title change is editorial and is done for consistency.

CTS Section 3.11.2.1 provides limitations in offsite dose rate from radioactive material released in gaseous effluents from the site. In Subsection 3.11.2.1.b, the licensee i

proposed to delete lodine - 131, lodine - 133". CTS Section 3/4.11.2.3 provides limitations for gaseous effluents and doses from radiciodines, radioactive materialin '

particulate form and radionuclides other than noble gases. The licensee proposed to delete "RADIOIODINES" in the title of the specification and the bases, and to delete " lodine - 131, '

lodine - 133". The licensee stated that the plant was shut down on July 22,1996.

Except for 1-125 (half-life - 59.5 days),1-129 (half-life - 1.6E7 years), and Kr-85 (half-life

- 10.8 years), the spent fuelinventory of the dose-contributing radioactive iodine and noble gas isotopes has decayed more than 20 half-lives since shutdown (i.e., less than 0.0001 %

of the original amount remainsi. In addition, the definition for " Dose Equivalent 1-131" l

(" Standard Technical Specifications, Westinghouse Plants," NUREG-1431) does not include  !

l-125 and 1-129 in the dose assessment due to their negligible inventory in the spent fuel. l Except for Kr-85, the noble gas nuclides that contribute to a whole body dose have also decayed to a negligible amount. CYAPCO has performed fuel handling and cask drop i

accident dose calculations, which conclude that doses (i.e., whole body and thyroid) at the l Exclusion Area Boundary are a small fraction of the 10 CFR 100 dose limits and the EPA PAGs. l The limitations in these specifications were established to assure that all releases would be \

1 within the dose limits specified in 10 CFR Part 20. With the reactor permanently defueled, radioactive gases are no longer generated and there is no longer a need to degasify primary coolant. The staff has determined that due to the permanently shutdown and defueled status of the Connecticut Yankee facility, the changes to this section are appropriate, due to the decay of short-lived radioactive isotopes in the spent fuel and are acceptable.

CTS Section 5.0 Section 5 describes the design features of the Connecticut Yankee facility. In accordance l with 10 CFR 50.36(c)(4), this section is intended to describe features of the facility such as j

materials of construction or geometric arrangement that,if altered, would significantly affect safety and are not covered in other sections of the technical specifications. The CTS contain the design features in seven areas: 5.1

• Site," 5.2 " Containment," 5.3

" Reactor Core," 5.4 " Reactor Coolant System," 5.5 " Meteorological Tower Location," 5.6

" Fuel Storage.~and 5.7 " Reactor Vessel Design Transients." The licensee is proposing to retain Sections 5.1 and 5.5 without changes. These subsections reflect systems that are l l

l retained in Technical Specification Sections 3/4.

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Sections 5.2 - Containment, 5.3- Reactor Core, 5.4- Reactor Coolant System, and 5.7 -

Reactor Vessel Design Transients were deleted, including associated tables. The deleted subsections and their associated tables address those design features associated with fuel in the reactor vessel or in the containment. The staff has determined that due to the permanently shutdown and defueled status o' the Connecticut Yankee facility, these subsections are no longer necessary for safe operation or maintenance of the plant, and finds the deletions to be acceptable.

In Section 5.6, Fuel Storage, the licensee proposed to change the title from " FUEL STORAGE" to " SPENT FUEL STORAGE"; change ~5.6.1 CRITICALITY" to " CRITICALITY",

renumber subsection ~5.6.1.1" to ~5.6.1"; delete "NEW FUEL"; delete Subsection 5.6.1.2:

change Page ~5-5" to Page ~5-4"; delete Figures 5.6-1 and 5.6-2; and, move the remainder of Section 5.6 from Page 5-Sa to the new Page 5-4. The changes to Section 5.6 was proposed, since the new fuel was expected to be removed from the site prior to the issuance of this proposed amendment. The staff verified that new fuel was removed from the site in 1997 (reference NRC Inspection 97-06). The changes to Section 5.6 reflect the removal of the new fuel from the site and are acceptable.

CTS Section 6.0 The licensee proposed adding item 6.8.1.g, Spent Fuel Pool Cooling and Monitoring Program to the TS. The Spent Fuel Pool Cooling and Makeup Monitoring Program requires that the primary methoe br spent fuel pool cooling and the primary method for spent fuel pool makeup capabilit) be monitored and maintained. The licensee states that the SFP cooling system will have available two cooling pumps and two heat exchangers, and that each heat exchanger is capable of removing the decay heat load. The program will provide reasonable assurance that the equipment, components, systems and water sources used for spent fuel pool cooling and for maintaining spent fuel pool water level are capable of fulfilling their intended functions and are protected against freezing. Due to the permanently shut down and defueled status of the plant, the staff finds this change acceptable.

3.0 STATE CONSULTATION

in accordance with the regulations of the Commission, the of State of Connecticut, Department of Environmental Protection, official was notified of the proposed issuance of this amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and also changes a reporting requirement. The amendment also relates to administrative procedures or requirements. The NRC' staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents 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 the amendment involves no significant hazards consideration, and there has been no public comment on such finding (62 FR 63978). Accordingly, the amendment meets the eligibility criteria for categorical exclusion sct forth in 10 CFR 51.22(c)(9) and (10).

Pursuant to 10 CFR 51.22(b) no environmentalimpact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

5.0 CONCLUSION

The Commission 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 regulations of the Commission, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. '

Principal Contributors: W. Raymond T. Fredrichs Date: June 30, 1998 x

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