B15252, Application for Amend to License DPR-65.Amend Would Increase Allowable Post Accident in-leakage Rate for CR

Application for Amend to License DPR-65.Amend Would Increase Allowable Post Accident in-leakage Rate for CR

ML20086G125
Person / Time
Site:	Millstone
Issue date:	07/07/1995
From:	Debarba E, Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES SERVICE CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20086G129	List: B15252, Application for Amend to License DPR-65.Amend Would Increase Allowable Post Accident in-leakage Rate for CR ML20086G132 ML20086G144 ML20086G153 ML20086G163 ML20086G170
References
B15252, NUDOCS 9507140140
Download: ML20086G125 (8)

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\ [ E 8IOS h 80ER Northeast Utuities Service Comi>any a' P.O. Box 270 4 Ilartford, CI' 06141-0270 F (203) 665-5000 July 7, 1995 Docket No. 50-336 ,

B15252 l

Re: 10CFR50.90 b

U.S. Nuclear Regulatory Commission p Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Control Room Emeroency Ventilation System Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend Operating License DPR-65 by incorporating the changes identified in the Attachments into the Technical Specifications of Millstone Unit No. 2. The proposed changes to the Technical Specifications will increase the allowable post ,

accident in-leakage rate for the control room, provide revised differentia 1' pressure values across filters, and incorporate an editorial enhancement to the Technical Specifications. The

) proposed marked-up pages of the existing Technical Specifications l

are provided in Attachment 1. The retyped pages are provided in Attachment 2.

Descrintion of the Pronosed Chances The proposed change to technical specification 3/4.7.6 is being made to: 1) increase the allowable control room air conditioning (CRAC) system in-leakage from 100 cubic feet per minute (cfm) to 130 cfm; 2) provide a more conservative value for the maximum differential pressure across the high efficiency particulate air (HEPA) filters and charcoal adsorbers; 3) clarify that when the CRAC system is shifted to " recirculation," this will be performed from the normal mode; and 4) modify the corresponding basis to reflect the above changes and to note that there are certain infrequent situations during which the control room emergency ventilation system (CREVS) will not automatically operate.

Safety Assesst4ED1 The proposed license amendment will address two technical modifications. The first proposed modification will modify the acceptance criterion for the differential pressure across the HEPA filters and the charcoal adsorbers from its current value of 6 oc<n uv s.%

950714o140 950707 PDR ADOCK 05000336 P PDR {g j 4

d U.S. Nuclear Regulatory Commission B15252/Page 2 July 7, 1995 inches water gauge to its new value of 3.4 inches water gauge. The second proposed modification will increase the allowable post accident in-leakage rate for the control room from its current value of 100 cfm to its new proposed value of 130 cfm. The proposed change will also incorporate an editorial enhancement which will clarify that the control room ventilation system must be I operating in the norma mode when the system is tested to verify that it switches on a 4ocirculation signal into a recirculation mode.

The CRAC system consists of two full capacity, completely independent air handling and mechanized refrigeration subsystems.

The CRAC system has the capability of ventilating with outside air or cooling, using mechanical refrigeration during normal, shutdown and accident conditions.

During normal operation, the CRAC system is operated in the l

" normal" mode. In this mode, one CRAC supply fan and one CRAC exhaust fan are required to be in operation. Outside air is introduced into the system, at a rate of approximately 400 to 800 cfm, through the outside air supply louvers. This outside air is supplied through two air dampers to the suction plenum of the CRAC system supply fans. In the " normal" mode, the control room filtration subsystem is in standby status with the filter fans secured and the discharge dampers closed.

The CRAC " recirculation" mode will be automatically initiated by any one of the following conditions:

1. Auxiliary Exhaust Actuation Signal

2. Enclosure Building Filtration Actuation Signal

3. Intake Duct High Radiation Condition

4. Loss of Power to the Radiation Monitors in the Control Room Outside Air Supply.

An automatic initiation of the CREVS will shift the CRAC to the recirculation mode of operation in which outside air is not introduced into the system and the flow of air to the cable vault and the outside are isolated. One supply fan and one exhaust fan will remain in operation with approximately 15,000 cfm of air recirculating through the system and approximately 2,500 cfm of the air flow passing through the control room filtration subsystem, l which consists of particulate, HEPA and charcoal filters and fans.

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The radiological dose calculations use the nominal design i recirculation flow rate of 2,500 cfm. (It should be noted that l

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U.S. Nuclear Regulatory Commission B15252/Page 3 July 7, 1995 2,500 cfm is also the typical operating flow based on past history.) This is acceptable because the effects on the 1 calculation of the ranges of flows that are possible are I overwhelmed by other conservatisms used in calculations.

The modification to surveillance requirement 4. 7. 6.1 will revise the acceptance criterion for the differential pressure value across the HEPA filters and charcoal adsorbers and will increase the amount of in-leakage for the control room to 130 cfm.

These proposed changes have been included in the revised control room habitability calculations for Millstone Unit No. 2. The results of these calculations show that the resultant doses that will be seen by the operators are at or near current levels, and continue to be within regulatory limits.

l l The accidents which have been reviewed for their impact on the l Millstone Unit No. 2 control room are the Millstone Unit No. 1 main I

steam line break (MSLB), the Millstone Unit No. 2 loss of coolant accident (LOCA) low wind speed case and high wind speed case, and the Millstone Unit No. 3 LOCA. The Millstone Unit No. 1 LOCA was also reviewed but the results are bounded by the Millstone Unit No. 2 LOCA. The assumptions that were used in the calculations are summarized and contained in Attachment 3. These assumptions are different than those used for the calculation results found in the Final Safety Analysis Report. These new assumptions are considered reasonable and appropriate.

The najor change with the new assumptions is that we now rely on operator actions in certain situations, to properly align the system. This has been reflected in the Bases for Technical Specification 3/4.7.6. Although the system will more than likely be configured automatically within 42 seconds, there are some low l probability events that will impact the single failure proof l capabilities of the system for which operator action will be needed to properly align the fans. The increase in the allowable in-leakage value in and by itself would result in higher control room doses. However, this has been offset by the use of more realistic assumptions in the accident analysis. The main contributor in the equalizing of the old and new control room doses is the use of the iodine dose conversion factors found in ICRP 30.

This guide has been used by the NRC in other applications for the past five years. Use of this standard results in more. realistic doses which are approximately 30 percent lower than calculated earlier. The assumptions used in the new calculations are presented in Attachment 3, Table 1. Using the revised assumptions, the resultant offsite doses are close to the original values and in all cases are within regulatory limits. The resultant doses are contained in Attachment 3, Table 2. A copy of each of the four

U.S. Nuclear Regulatory Commission B15252/Page 4 July 7, 1995 calculations are also contained in Attachment 3. The proposed modification will also clarify Surveillance Requirement 4.7.6.1.e.2 and the associated bases. The function of the smoke detectors in the return ductwork will not be affected by these proposed modification. The smoke detectors will continue to isolate the supply unit and initiate purging cperations in the event of a smoke condition. This function will continue to be done automatically without operator intervention.

The modification to the acceptance criterion for the HEPA filters and charcoal adsorbers will result in the use of a plant specific value in lieu of the current value, which is a hold over from the Combustion Engineering Standard Technical Specifications. The 6 inch value should have been replaced with a plant specific value when the current version of the technical specifications were written. This proposed modification will replace the 6 inches water gauge value with a plant specific value of 3.4 inches water gauge.

The proposed modification will also clarify Surveillance Requirement 4.7.6.1.e.2 and the associated bases section. Since the proposed changes do not significantly modify the dose in the control room, and are within the regulatory limits, the proposed changes are considered safe.

SIGNIFICANT HAZARDS CONSIDERATION NNECO has reviewed the proposed changes in accordance with 10CFR50.92 and has concluded that the changes do not involve a significant hazards consideration (SHC). The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed changes do not involve an SHC because the changes will not:

1. Involve a significant increase in the probability or consequences of an accident previously analyzed.

The CRAC system in the recirculation mode is used to mitigate the effects of an accident. Surveillance Requirement 4.7.6.1.e.2 has been modified to clarify that the system will automatically switch from the normal mode into a recirculation mode. This change and the proposed modifications to the acceptance criterion for the differential pressure across the HEPA filter 3 and charcoal adsorbers and the increase in the control room in-leakage have no affect on the probability of an accident previously evaluated. The consequences of the accidents that have been previously evaluated have been reviewed to determine the impact of these proposed modifications. The increase in the in-leakage will affect the

U.S. Nuclear Regulatory Commission >

B15252/Page 5 July 7, 1995 results of previously generated accident analysis. The accidents evaluated, namely the Millstone Unit No. 1 MSLB and '

LOCA, Millstone Unit No. 2 LOCA, both high and low wind speed case, and Millstone Unit No. 3 LOCA have been reviewed. The i Millstone Unit No. 1 LOCA doses to the Millstone Unit No. 2  ;

control room were qualitatively determined to be bounded by the Millstone Unit No. 2 LOCA cases. Therefore the Millstone Unit No. 1 LOCA was not performed. The remaining accidents were performed. The resultant doses are nearly identical to the existing doses found in the Millstone Unit No. 2 Final Safety Analysis Report and are all within the regulatory limits. To perform these revised control room dose calculations, NNECO used certain new assumptions which NNECO ,

believes better model the control room and the effects the accident will have on the control room. The most significant change with the assumptions is the use of ICRP 30 in lieu of Regulatory Guide 1.109, Revision 1 for iodine dose conversion factors. The NRC has used ICRP 30 over the past 5 years for other applications and its use in this instance is appropriate.  ;

The change in the acceptance criterion for the differential pressure across the HEPA filter and charcoal adsorbers is a conservative modification in that the value given is a plant specific value and will be more indicative of blocked or clogged filters in actual plant conditions. These proposed changes do not have any negative impact on the consequences of any accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

The proposed modifications to Surveillance Requirement 4.7.6.1 will clarify a portion of a surveillance requirement and will modify the differential pressure across the HEPA filters and the charcoal adsorbers. These changes will not create the ,

possibility of a new or different kind of accident from any previously evaluated. The increase in the allowable control .

room in-leakage value from it current level of 100 cfm to its  !

new value of 130 cfm also does not create the possibility of l a new or different kind of accident. The CRAC system is used to mitigate the consequences of an accident.

3. Involve a significant reduction in the margin of safety.

The proposed modifications do not decrease the margin of safety provided. Using the new accident assumptions, the limiting accidents were re-calculated to determine the impact r on the Millstone Unit No. 2 control room. These values are

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U.S. Nuclear Regulatory Commission B15252/Page 6 July 7, 1995 similar to the values found in the Millstone Unit No. 2 Final Safety Analysis Report and the Millstone Unit No. 2 Safety Evaluation Report and are within the regulatory limits established for the control room operators. Since the re--

calculated doses have been shown to be within limits, it has been concluded that there is no reduction in the margin of safety.

Moreover, the Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (51FR7751, March 6, 1986) of amendments that are considered not likely to involve an SHC. The change to Surveillance Requirement 4.7.6.1 and its Bases are not enveloped by the cited examples but nonetheless have been shown to not constitute an SHC.

Envi ronmenta1 Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed changes do not increase the types and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concluded that the proposed changes meet the criteria delineated in 10CFR51.22 (c) (9) for a categorical exclusion from the requirements for an environmental impact statement.

Nuclear Safety Assessment Board The Nuclear Safety Assessment Board has reviewed and approved the proposed changes and has concurred with the above determination.

I State Notification In accordance with 10CFR50. 91 (b) , we are providing the State of Connecticut with a copy of this proposed amendment to ensure their awareness of this request.

Schedule Regarding our proposed schedule for this amendment, we request issuance at your earliest convenience with the amendment effective as of the date of issuance, to be implemented within 60 days of issuance.

Conclusion The proposed changes have been reviewed in accordance with 10CFR50.92 and have been determined to not constitute an SHC. In

U.S. Nuclear Regulatory Commission B15252/Page 7 July 7, 1995 addition, the proposed changes have been reviewed against 10CFR51.21 and it has been determined that the proposed changes meet the criteria for a categorical exemption for an environmental impact statement.

The marked-up pages of the existing Technical Specifications are provided in Attachment 1. The retyped pages are provided in Attachment 2 and reflect the currently issued pages.

If you should have any questions on the above or attached, please contact Mr. M. Robles, Jr. at (203) 440-2073.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY FOR: J. F. Opeka Executive Vice President BY: E. A. DeBarba Vice President cc: T. T. Martin, Region I Administrator G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 P. D. Swetland, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 Mr. Kevin T.A. McCarthy, Director Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street P.O. Box 5066 Hartford, CT 06102-5066 Subscribed and sworn to before me this day of Oahu ,

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Docket No. 50-336 B15252

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Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Control Room Emergency Ventilation System Marked Up Pages i

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