ML20058Q270

Forwards Draft Document from Licensee Re Proposed Rev to TS Suppl Leak Collection & Release Sys

ML20058Q270
Person / Time
Site:	Millstone
Issue date:	10/21/1993
From:	Rooney V NRC
To:	NRC
References
NUDOCS 9310260123
Download: ML20058Q270 (39)

Text

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October 21, 1993 -

Millstone 3, 50-423 NOTE TO: Docket File 50 - 423 p FROM: V. L. Rooney / ~ k(d_ .,.

BUBJECT: Draft Received 1

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Attached is a draft document that was faxed, and reached my desk I 10/23/93.

Attachments: As stated.

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DRAF^T3 October XX, 1993 Docket No. 50-423 l B14647 Re: 10CFR50.90 10CFR50.91  ;

U.S. Nuclear Regulatory Comission Attention: Document Control Desk Washington, DC 20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications '

Sucolementary leak Coll _ection and Release System Introduction Pursuant to 10CFR50.90, Northeast Nuclear Energy Corporation (NNECO)' hereby proposes to amend its Operating. License, NPF-49, by incorporating the changes identified in Attachments 1 and 2 into the Technical Specifications of Millstone Unit No. 3. The markup technical specification pages'are provided in Attachment 1, and the retyped technical specification pages are provided in ,

Attachment 2. The proposed changes to Millstone Unit No. 3 Technical Specifications '4.6.6.1.d.3 and 3.6.1.2.a will increase the time allowed to achieve a negative pressure of 0.25 inches water gauge within the secondary containment boundary from 50 seconds to 150 seconds,.and reduce the allowable  !

integrated containment leakage rate (L.) from 0.65 wt.% to 0.30 wt.5 of the .

containment Also, NNECOair per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is proposing at theBases to revise design basis p/4.6.6.1 Section 3 by addingressure, an respecti:

additional- discussion regarding the time requirement- of Technical Specification 4.6.6.1.d.3. In addition, NNECO is' requesting that the NRC ,

Staff process this license amendment request on an emergency basis pursuant to  :

10CFR50.g1(a)(5). Emergency authorization is required by October XX, 1993, to ,

Mt11 stone Unit No. 3 is shutdown (Mode In 5)).parallel supportwithstart up effort, this (i.e., entry the into Mode 4 NRC Staff may wish to consider.whether it is advisable to exercise enforcement discretion from Technien1 Specifications 3.6.6.1, 3.6.1.2.a, and 3.7.9 to be  ;

effective until the license amendment is issued. The enforcement discretion-would permit NNECO to start up and operate Millstone Unit No. 3 while the

-proposed license amendment. is being processed. NNECO believes that an .. g onergency license amendment is warranted in this case to pemit the start-up i and operation of the plant, since the associated operational risk with .the I request has no negative impact.on public health and safety.

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DRAFT /U.S. Nuclear Regulatory Commission  ; L 4 g B14647/Page 2 October XX, 1993 In the near future, NNECO plans to submit an additional proposed license amendment which further revises the Millstone Unit No. 3 Technical Specifications regarding this issue. The subsequent submittal will propose to reinstato the current upper bound for the overall integrated leakage rate of 0.65 wt.'t, of the containment air per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period at design basis pressure, and to increase the time required to draw a negative pressure of 0.25 inches water gauge within-the secondary containment boundary to approximately 5 minutes.

Systen Descriotions The operability of the supplementary leak collection and release system (SLCRS) ensures that radioactive materials that leak from the primary containment into the secondary containment boundary following a design basis accident (DBA) are filtered out and adsorbed prior to any air release to the environcent. The SLCRS system is a two-train filtration system with an independent duct system. Currently, the design requirement for the SLCRS is to achieve a negative pressure of 0.25 inches water gauge within the secondary containment boundary within one minute of a DBA. The secondary containment boundary is comprised of the containment enclosure building and contiguous buildings (main steam valve building [ partially), engineered safety features building [ partially), hydrogen recombiner building [ partially), and auxiliar{

building). The secondary containment boundary is referred to as the ' annulus in Millstone Unit No. 3 Technical Specification 4.6.6.1.d.3.

In a proposed license amendment dated July 29, 1993,"3 NNECO sro osed changes to the Millstone Unit No. 3 Technical Specifications related ;o $he SLCRS and auxiliary building filtration system (ABFS). The principal purpose of the proposed cha is to address the phenomenon described in Information Notice 68-76'gges for Millstone Unit No. 3. The proposed changes clarify the location within the secondary containment boundary where a negative pressure of 0.25 inches water gauge must be obtaineds delineate the equipment required to comprise an operable SLCRS and denote the equipment required to comprise an operable ABFS. Also, NNECO proposed to replace the term " annulus" denoted in Technical Specification 4.6.6.1.d.3 with the phrase " secondary containment boundary." The phrase " secondary containment boundary" was defined in Bases Section3/4.6.6.1. These changes would permit Millstone Unit No. 3 to operate with the potential for the upper elevations of the enclosure building to be at a slightly positive pressure. In that submittal, MNECO requested that the NRC (1) J. F. Opeka letter to the U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications, Supplementary Leak Collection and Release System," dated July 29, 1993.

(2) C. E. Rossi letter to All Holders of Ooerating Licenses or Construction Permits for Nuclear Power Reactors, "dRC Information Notice No. 88-76:

Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control,' dated September 19, 1968.

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DRAFT /U.S. Nuclear Regulatory Commission B14647/Page 3 N j 8 A W []

October XX, 1993 Staff issue the subject amendment prior to October 31, 1993. We hereby reaffirn our request for amendment issuance on or before that date.

The auxiliary building ventilation system (ABVS), which includes the ABFS, provides the normal and postaccident means for cooling of vital equipment located in the auxiliary building. Additionally, the system augments the SLCRS in its design basis function of drawing a negative pressure of 0.25 inches water gauge Twithin the secondary containment boundary. One train of the cooling portion of this system operates normally in support of plant operation. The ABFS portion of the system is normally in a standby mede except for occasions when the system is operated manually to support the removal and filtration of radioactive gases released while drawing reactor coolant system samples, and during monthly charcoal filter bed testing. In the event of a DBA. one train of the ABVS is automatically brought into operation to support the SLCRS and to provide cooling of vital equipment.

The design of the ABFS incorporates two redundant trains of active equipment which share common ductwork and plenums (see Attachment 3 for a simplified diagram). Each train of the ABVS is powered via redundant and independent power supplies.

There are no normal or accident modes of o >eration wherein it is acceptable to automatically and simultaneously operate both trains of the ABVS due to air flow limitations on the common ductwork. This system characteristic (single train operation under normal and accident conditions) requires the inclusion of a design feature wherein the idle train of the ABVS must first verify, through the measurement of process (air) flow, that the preaccident operating train of equipment is positively shutdown before automatically starting. This is unlike most safety systems wherein both trains of equipment automatically start and run independent of one another. This design feature results in delayed starting of the standby train of equipment, thereby adding time to the process of drawing a negativo pressure of 0.25 inches water gauge within the secondary containment boundary. It is this feature that precludes the ability to declare the ABVS trains not only redundant, but also independent.

Backaround/Secuence of Events 1992 Events On September 29, 1992, the 'B' train of the SLCRS was declared inoperable, and it was determined that insufficient surveillance testing existed to prove the operability of the "A" train. Specifically, timing delays in the ABVS fan circuitry resulted in a 70 75 second delay in ABVS fan start from signal actuation. The ABVS system acts in parallel with the StCRS and can affect the ability of the SLCRS to draw a negative pressure of 0.25 inches water gauge in the secondary containment boundary. In addition, NNECO determined that the ,

timing sequence difference between an actual accident configuration and the existing 3LCR$ drawdown surveillance was large enough to consider the surveillance inadequate for verifying system operability.  :

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DRAFT /U.S. Nuclear Regulatory Comission i t g

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B14647/Page 4 October XX, 1993 The. immediate corrective action based on Limiting Condition of Operation (LCO) 4.0.3 was to perform another in-service test (IST) to determine the operability of the "A" train of SLCRS.

While performing the second IST cn September 30, 1992, the "A" train of the SLcRS failed to draw down the secondary containment boundary within the required time frame And was declared inoperable.

the 0.25-inches negative pressure criterionnot could(The be met in IST results shcKed tha the rquired lhe 60 shutdown seconds [80 5seconds to Mode actual])d on October 1.1992.NNECO began a shutdown of the was complete In a letter dated November 12, 1992,* NNECO described the background, status, and course of action taken for the resolution of the design deficiencies related to the ABVS and the SLCRS. During the month of October 1992, NNECO completed several modifications pricr to the start-up of the plant.

Current Refueling outage During the current (cycle 4) refueling outage at Hillstone Unit No. 3. the 18-month surveillance testing of the ABFS train "A" fan identified a delay in the start caused by an inherent design characteristic of the flow switch in the circuit. The flow switch is physically located at the train "B" fan to sense train "B" air flow. Due to the design of the flow switch, the train "A" fan would not be permitted to start imediately after a loss of power (LOP). The train "A" ABFS fan started approximately 35 seconds late after receiving a sequenced safeguards signal during the LOP testing. This starting delay was repeated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later during the engineered safety features / loss of power (ESF/ LOP) test.

In the event of an ESF actuation (i.e., upon a safety injection signal [ SIS))

with a LOP, the ABFS also provides an exhaust path to assist the SLCRS in drawing a negative pressure of 0.25 inches water gauge in the secondary containment boundary. Based on previous test results, the approximate 35 second time delay for starting the ABFS train "A" fan means that train "A" of the SLCRS would not draw a negative pressure of 0.25 inches water gauge in the secondary containment boundary within the 60 second requirement. Instead, it is projected that the required negative pressure would not be reached until approximately 70-80 seconds. At 60 seconds, the secondary conta.inment boundary would have reached a negative pressure between 0.10 and 0.20 inches water gauge. A similar flow switch was functionally deleted from the "B" train of the ABF3 in October 1992.

A full investigation to determine the impact of the flow switch design on other safety related ventilation systems was undertaken. In Licensee Event (3) J. F. Opeka letter to the U.S. Nuclear Regulatory Comission,

" Supplementary Leak Collection and Release System / Auxiliary Building Ventilation System-Event Sumary," dated November 12, 1992.

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performance of this test, the "B" train ABFS fan did not start until 90 seconds after a SIS. This failure rendered the SLCRS train inoperable.

Task Force NNECO's review of this failure has identified additional single failure vulnerabilities within the SLCRS instrumentation and controls. To resolve this issue, a team of engineers was assembled to address the cause of the deficiencies in the system. The matter has been pursued on a seven-day per week, extended hour basis as a top corporate priority. The team consists of '

representatives from the Probabilistic Risk Assessment Group, the Project >

Services Department, the Engineering Department, Plant Engineering, and '

Huclear Licensing.

Design changes The design changes described below are bein0 impleanted and will allow  ;

specific equipment to start as soon as aossible after a LOP event coincident with an accident signal to ensure that SLCRS operating in conjunction with the ABVS will achieve drawdown of the secondary containment boundary to a negative pressure of 0.25 inches water gauge within 2.5 minutes following an accident signal; i.e., LOP, SIS, containment depressurization actuation signal (CDA). ,

1. ABFS Exhaust Fan 3HVR*FN6B Chances Under the present logic, a LOP coincident with the SIS /ESF signal is required in series with a high plenum pressure to initiate a 30 second time delay before opening fan 6B's inlet and outlet dampers. These dampers are required to open before fan 6B can start. The purpose of this circuit is to detect fan 6A failure and to start the other train ,

i auxiliary building exhaust fan. A time delay in this circuit is required to allow fan 6A to start after a LOP, and thus satisfy the pressure switch in the plenum. However, the time delay (i.e., the window of opportunity for fan 6A to start) should begin at the time of the event. The logic change starts this 30-second timo delay at the initiation of the LOP / SIS /ESF signal. Previous testing has demonstrated that fan 6A has started and satisfied the plenum pressure switch prior to the 30-second time period.

In addition, a reset feature will be added to the Spec. 200 Microprocessor Logic using fan 68 control switch 1A-3HVR*FN6B located at main ventilation panel (3HYS*PNLVP1). This feature will ensure that in (4) 5. E. Scace letter to the U.S. Nuclear Regulatory Comission, " Facility Operating License No. NPF-49, Docket No. 50-423, Licensee Event Report 93-014-00," dated September 30, 1993.

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October XX, 1993  :

order to shut down fan 6B after an accident condition, two deliberate 4 operator- actions must be perfor:nedt i.e., reset safety signal LOP, SIS, CDA and place 3HVR*FN6B control switch in stcp. -

2. Charoina pumo and Reactor Plant Comoonent Coolina Water (RpCCW) Suon1v Fan 3HVR*FN14A/B amt Fvhaust Fan 3HVR*FN13A/B Timing. Relav SetDoint Chances _

The proposed change to the timing relay setpoint from 20 seconds to-10 seconds will allow the opposite train fans to start earlier on-detection of low flow when a fan failure occurs. This change will reduce the time delay experienced for starting fan 6B during a test performed under train "A" failure (train "A" emergency diesel generator _

failure) coincident with an accident signal and LOP. Early starting of  ;

the opposite train fans will result in faster planum pressure increase which will initiate starting of fan 6B faster. Also, the revised time "

delay will maintain adequate margin of time for the primary fan to start and clear the low flow condition before the opposite train fan can start. .

The operation procedure change for the fan control switch in AUTO _!

position will climinate the potential for operating both train fans simultaneously due to a flow switch failure or a fan failure. Operating both train fans could adversely affect the flow balance with ABFS operation during an accident.

3. Charaino Purro and RPCCF Area Sucoly Fans 3HVR*FN14A/B and Exhaust Fans -

3HVR*fN13A/B Flow Swite1 Setooint chpnces The flow switch setpoint changes will allow the flow switches to respond without excessive time delay to the low flow condition resulted by a fan failure and yet maintain adequate nargin not to cause a spurious low flow signal when the fan operates normally. This change will reduce the excessive time delay experienced for starting fan _68 during the IST. j i

System Testing Program and Results j l

' 1 A performance assurance program has been implemented to perform an intensive I assessment to demonstrate that the ABVS will operate to meet its design - ll intent. It consists of a combination of tests and/or analyses to demonstrate i system response to safety injection signals under both normal and. failure I conditions in the ABV5 or its power supply systems.

Testing of individual components critical to correct operation of the system has been completed to assure -the development of time lines for equipment operation. Further, tests are being conducted during the week of October 18, 1993, to demonstrate system capability and operability.

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October XX, 1993 Additional tests are being developed for conditions where a failure occurs in a power supply system. Where the failure is within the ABVS. analysis will be used to evaluate system performance, except for certain specific failures.

Each operational mode of the ABVS will be considered and time lines for equipmer.t operation will be developed. These will be verified by test for normal and failure modes. Each time lino configuration will be tested, except where time lines for one test are the same as for another test or are bounded by tests for other time lines. Documentation supporting failuro analysis conclusions will be prepared and such analysis will be traceable to tests documenting the validity of the analogies.

Perfomance assurance tests will be based on prior evaluation of system response to the tested conditions and which evaluations support acceptable operation. Certain tests will take place in conjunction with SLCRS operation to demonstrate the capability to obtain a negative pressure of 0.25 inches of water gauge within the required time. Tests will be traceable to these bounding tests which demonstrate system operability.

Descriotion of Procosed Chances NNECO proposes to revise the Millstone Unit No. 3 Technical Specifications by reducing the upper bound of the overall integrated leakage rate required by Technical Specification 3.6.1.2.a from 0.65 wt.% to 0.30 wt.% of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at design basis pressure, by increasing the time required by Technical Specification 4.6.6.1.d.3 to draw a negative pressure of 0.25 inches water gauge within the secondary containment boundary from 50 seconds to 150 seconds (this time does include the diesel generator start and load time of approximately 10 seconds), and by revising Bases Sectier 3/4.6.6.1 to provide a more detailed explanation for the time required by Technical Specification 4.6.6.1.d 3 to draw a negative pressure of o.a5 inches water gauge within the secondary containment boundary.

NNECO proposes to revise Technical Specification 3.6.1.2.a by reducing the upper bound of the overall integrated leakage rate required by Technical Specification 3.6.1.2.a from 0.55 wt.% to 0.30 wt.% of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at design basis pressure. 1his proposed change is more restrictive than the current Technical Specification requirement. This revision to the Technical Specifications has been proposed to enable NNECO to also revise Technical Specification 4.6.6.1.d.3. ~

NNECO proposes to revise Technical Specification 4.6.6.1.d.3 by increasing the time required to s' raw a negative pressure of 0.25 inches water gauge within the secondary containment boundary from 50 seconds to 150 seconds (this time does include the diesel generator start and load time of approximately 10 seconds). In addition NNECD proposes to revise Bases Section 3/4.6.6.1 by adding an additional discussion concerning the time requirement of Technical Specification 4.6.6.1.d.3.

The proposed changes to the Techt.ical Specifications presented in Attachments 1 and 2 reflect the currently issued version of the Millstone Unit No. 3 l

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DRAti/U.S. Nuclear Regulatory Commission 3RA54gna B14647/Page 8 October XX, 1993 Technical Spectfications. The proposed , changes to the Technical Specifications submitted on July 29, 1933, are not reflected in the enclosed retype. These proposed changes are discussed in detail in the System Description Section of this letter. In the July 29, 1993, submittal, NNECO requested that the NRC Staff 1ssue the subject amendment prior to October 31, 1993.

In the near futures NNECO plans to submit an additional proposed license ,

amendment to the Millstone Unit No. 3 Technical Specifications regarding this issue. The future submittal will propose to reinstate the current upper bound for the overall integrated leakage rate of 0.65 wt.% of the containment air per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period at design basis pressure, and to increase the time required to draw a negative pressure of 0.25 inches water gauge within the secondary containment boundary to approximately 5 minutes.

Safety Assessment The ability of the combined SLCRS/ABVS to meet the proposed Technical Specification requirement to draw a negative pressure of 0.25 inches water gauge in the secondary containment boundary within 150 seconds of a start signal is established through the evaluation of modification-related operating time changes and the use of prior test data. These tests show that the SLCRS/ABVS equipment is capable of developing well in excess of 0.25 inches water gauge within the seccndary contair. ment boundary. There is reasonable assurance that this can be accomplished within the 2.5 minute time period.

Furthermore. testing currently underway will be completed following implementation of previously -described modifications. This testing will validate the system's ability to perform its intended function in the requisite time frame. The testing program and any resulting modif tcations will provide reasonable assurance that the systems are operable, and that i single failure vulnerabilities have been satisfactorily addressed.

To further assure operability, a failure modes and effects analysis (FMEA) of ,

the ABVS using the probabilistic safety study (PSS) arecess was conducted to '

determine if there are any single failure vulnerabilities. ABVS components I whose operation could potential,y affect the SLCRS performance were included l in this analysis. Special emphasis was placed on those components which .

brou0 h t about interactions between the two trains of ABVS. The ABVS l fans 13A/B,14A/B, 6A/B, the flow switches FS27B, FS52A/B, FSSBA/B are some of i the si nificant 0 components that were included in the FMEA. Since SLCRS is a standby system that is needed subsequent to a release into the secondary containment enclosure, the analysis focused on SIS and LOP initiators.

i (5) J. F. Opeka letter to the U.S. Nuclear Regulatory Commission, "Millstor.e l Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specifications, Supplementary Leak Collection and Release System," dated  ;

July 29, 1993.

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October XX, 1993 Due to the complexity of the system, it was necessary to clearly identify a list of multiple failure criteria that would be used to determine the impact on the system. For example, in addition to a single fan failure to start, simultaneous operation of both trains of ABVS that could adversely affect the flow balance with ABFS operation during an accident was also considered.

During this FMEA, a suspected single failure (operation of all ABVS supply and exhaust fans) was confirmed and a design change was implemented to eliminate that single failure. A couple of other single failures that appeared to be failures on the wiring diagram were later determined to be not credible. In addition, the FMEA uncovered a potential for excessive pressurization of inlel plenum due to time delays built into the ABV5. Although system failure does not result due to this potential overpressurization, a design modification that involves a time delay will address the concern. No other credible active single failures of the ABVS were identified.

Extension of the time allowed to achieve drawdown of secondary containment from 1 minute to 2.5 minutes will have negligible impact on heating and cooling. Plant experience has shown that heatup and cooldown of thick-walled concrete structures, such as the Hillstone Unit No. 3 auxiliary building, is a relatively slow process. Following an accident signal, ventilation equipment is restarted promptly. Therefore, heatup or cooldown, during short periods while ventilation fans and/or heaters are inactive, is insignificant and can be neglected.

The proposed change to decrease the containment integrated leakage rate at the design basis pressure from 0.65 wt.%/ day to 0.3 wt.%/ day has been evaluated to determine the impact of the proposed lower leakage criterion on the Millstone Unit No. 3 containment test program. A review of the type "B" and 'C" leakage results for the current refueling outage shows that the total type "B' and "C' as-found and as-left respectively. leakage These are results were 0.096 significantly wt.%/

below thedaycurrent and 0.084 wt.%/dayl Technica Specification requirement of 0.39 wt.%/ day (0.6 L , when L, is equal to 0.65 wt.X/ day), and would be below the proposed limit of 0.18 wt.5/ day (0.6 L .

when L, would be equal to 0.30 wt.%/ day). Also, the results for bypass lpkage were 0.007 wt.%/ day for as-found and 0.008 wt.W/ day for as-left. These are below the current Technical Specification limit of 0.0273 wt.%/ day (0.042 L.,

when L, is equal to 0.65 wt.%/ day), and would be below the proposed Technical Specification limit of 0.0126 wt.%/ day (0.042 L., when L, would be equal to  !

0.30wt.%/ day). In addition, the results from the type 'A' test were 0.13268 l wt.%/ day for the as-found integrated leakage rate test and 0.13132 wt.%/ day '

for the as-left integrated leakage rate test. These are below the current Technical Specification limit of 0.4875 wt.%/ day (0.75 L,), and would be below the proposed Technical Specification limit of 0.215 wt.%/ day (o.75 L , when L, would be equal to 0.30 wt.%/ day).

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DRAFT /U.S. Nuclear Regulatory Commission B14647/Page 10 October XX, 1993 i NNECO has evaluated the above changes to determine the impact they would have  !

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on off-site overall effect doses during of the a design proposed basis loss changes wasofto coolant reduceaccident (LOCA) doses.

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The previously calculated exclusion area boundary (EAB) roid and whole body  ;

doses were 150 rem and 19.5 rem, respectively. Uti Izing the proposed. <

revisions, the EAB doses were calculated to be 146.5 rem thyroid and 9.45g rea '

whole body. The evaluation. concluded that the total curies of each iodine and noble gas isotope is~1ess over each time period for this analysis than for the.-

previous analysis. This indicates that the low population zone, . control room,-

and technical support center doses will be lower. Therefore, since .the  !

proposed changes result in a reduction in the calculated doses, they do not negatively impact public health and safety.  !

safety significance In the interest of providing a more global safety perspective regarding this  !

proposal, NNEC0 hereby provides the information contained in this sectiin. We  !

believe that this information is important because.it puts into perstective the risk significance of the SLCRS failure to meet its design requirement .

within the described time interval of 60 seconds. Given the conservatises in the design and analysis, the potential safety significance of not meeting the 60 second time interval is essentially zero because p_otential releases would -i be minimized as follows. Please note that our arososal contains an additional conservatism that is not considered. in the fol' ow' ng discussion (the proposed -

change to the upper bound for the overall integrated leakage rate).

• Using conservative DBA assumptions, a delay in the drawdcwn of the auxiliary building by the SLCRS to a negative pressure of 0.25. inches water gauge to a time up to 2 minutes would result in an unfiltered release tiat would remain below 10CFR100 dose limits; i.e., using Re ulatory Guide 1.10g dose factors, the calculated thyroid dose is ,

28 rem.

• If the same conditions are assumed, but ICRP 30 dose factors are ,

utilized, then the thyroid dose would be 180 ram. j

• Design basis LOCA radiological calculations are performed based upon some very simplistic, non-mechanistic assumptions. These simplifications were made because it was easier to fulfill the original purpose of the DBA radiological calculations which was to ensure reactor siting criteria could be met. This assurance was obtained through.the combined acceptability of plume dispersion to the EAB and low population-zone, containment leakage limits, and radioactivity reduction mechanisms.

such as containment sprays or secondary containment boundary filtration.

One of the simplifying non-mechanistic assumptions was:that gross . fuel failure resulting in 100 percent of the core noble gases and 50 percent of the core iodine inventory being released to the containment . air instantaneously at the time of the LOCA.

1 i

i

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October XX, 1993 Unfortunately, this simplifying non-mechanistic assumption was applied to the mechanistic desion requirements of the radioactivity reduction  ;

systems. This has resulted in some unnecessarily restrictive design  :

requirements for these systems. One example, applicable to thls

^

Technical Specification, is the requirement that in order to assume zero filter bypass, the secondary containment boundary be drawn to a negative i pressure.of 0.25 inches water gauge within one minute of the LOCA. ihe  !

one-minute requirement only exists because of the non-nechanistic  ;

assumption that core melt levels of radicactivity are instantaneously l leaking from the containment. Another non-mechanistic, unrealistic '

assunption is that until_ a negative pressure of 0.25 inches water gauge is achieved, all activity leaking into the secondary containment j boundary is immediately released to the environment unfiltered.

Both of the above two assumptions are physically impossible. Even in the absence of any emergency core coolant in.iection following a large break _ LOCA, it will take some time for the fuel elements to heat up a .

temperature at which significant release of fission products are expected. Per NUREG/CR 4881 ' Fission Product Release Characteristics i into containment under Design Basis and Severe Accident Conditions," l March 1988, the minimum time before release -of significant fission ,

products inventory is 6-10 minutes._ The simplistic source terms  ;

! developed for NUREG-1465, " Accident Source Terms for Light Water Nuclear l Power Plants" 1992 Draft, provide a more probable time for significant j fission product release for a pressurized water reactor-of 30 minutes, i A negative pressure of 0.25 incses water gauge could be established well  ;

within this time period. .;

I Although some of the fuel gap activity may be released at the onset of the LOCA, this is a small percent of the core inventory assumed to be i released in the DBA calculations. Hence, as long as the secondary containment boundary is sufficiently negative by the time we expect any i significant radioactivity leakage into the enclosure, there should be no j bypass leakage of any consequence.

• Additionally, any radioactivity which does leak into the secondary ,

containment boundary during the first 2.5 minutes of the event is not expected to be rolcased unfiltered. It is expected that under most  !

conditions, the secondary containment boundary will be . -0.25 inches  !

water gauge within one minute. For the potential scenarios where this j nay not be possible- (e.g., loss of normal power and failure of the ,

preferred diesel generator), the secondary containment boundary is still expected to be at negative pressure, although not -0.25 inches water

' gauge, within the approximate one-minuto time frame. Hence, there is no  ;

reason to believe any measurable fraction of the activity leaking into l the secondary containment boundary will meander to the outside walls and .

leak through any small openings.

Probabilistic Risk Assessment Insights Regarding the SLCR$ and ABFS

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DRAFT /U.S. Nuclear Regulatory Commission -

B14647/Page 12 3 October XX,1993 As a measure of the significance of this issue from a public risk perspective, note that the SLCRS and ABVS were not modeled in the Millstone Unit No. 3 PSS.

The reasoning is as follows:

• Public risk, defined as probability and consequences of accidents, is dominated by core melt accidents with gross containment failure (intersystem ,LOCA, gross containment isolation failure, containment t ruptureduetooverpressure)". The consequences for such accidents are in the range of 10' to < 10 man-rem.

• SLCRS and ABVS would be ineffective to mitigate such events because of the large magnitude of the release rate from the containment; i.e., they_

could not achieve a negative pressure. Given that a core melt with gross containment was because containment failure occurred,l systemsthe heat remova sprays most likely had also ) reason is that i failed.

In turn, this is most likely caused by loss o support systems such as all AC power. Therefore, SLCRS/ABVS are also likeTy failed for the risk dominant accident sequences.

• SLCRS and ARV5 do provide a benefit for core melt accidents with design leak rates. However, these are low consequence events to begin with.

For example, the Three Mile Island accident resulted in about 2000 man-rem dose to the public. For Millstone Unit No. 3, a consequence of approximately 1000 man-rem is used in the probabilistic risk assessment for core melt with design leakage rates and successful containment sprays. Much of that dose is from noble gases.

• SLCRS/ABVS would reduce releases of iodine activity by a factor of approximately 20, but the dose from noble gases, a significant '

contributor to the overall dose, is unaffected.

• Even though the Millstone Unit No. 3 PSS is a very detailed and extensive undertaking by industry standards, the subject systems were not even modeled in the PSS. This was because their inclusion would have had a negligible effect on the results of the study. Much of this stems from the significant differences between the hypothetical design requirements for these systems, versus the mechanistic and more realistic methodology employed in the PSS.

The PSS calculated risk to the public from core damage with subsequent containment leakage is only 2 person-rem over the remaining plant life (32 years). This is an insignificant fraction of the acceptable total ,

public risk from Millstone Unit No. 3.

Additionally, the calculated 2 sersen-rem assumes no credit for secondary filtration. Hence, th< s would be the maximum potential reduction in risk if all containment leakage (vice containment failure) was eliminated. SLCRS will not oliminate the noble gases which contribute most of the 2 person-rem. In addition, the first five l

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i DRAFT /U.S. Nuclear Regulatory Corrnission i B14647/Fage 13 '

October XX, 1993 minutes of release of any iodines, other volatiles, or particulates is  ;

calculated via probabilistic risk assessment techniques to be zero.

Thus, the more realistically calculated PSS public safety consequence or 1 the change is zero.  !

We request that the NRC consider the above perspectives when responding to l NNECO's request in this submittal. NNECO is fully committed to a conservative operating philosophf, and believes that the above considerations provide substantial evidence to that effect.

Justification for Emeroency License Amendment Pursuant , NNECO hereby requests NRC 5taff " emergency" approval of tothe 10CFR50.91[a)(5)dment propose: amen to its Operating License HPF-49.

Emergency authorization is required by October XX,1993, to allow start-up (entry lato Mode 4). At the present time. Hillstone Unit No. 3 is shutdown.

A discussion of the circumstances surrounding this situation and determination of the need for prompt action is provided in the Background / Sequence of Events Section of this letter and below.

First, it is important to recognize that the ABFS and SCLRS are highly interactive and by no means represent a conventional independent and redundant safety-related system. The intracracies of system design and their -

interrelationships have contributed significantly to the difficulties NNECO ,

has encountered in dealing with the issues discussed in this submittal. '

During the current (Cycle 4) refueling outage at Hillstone Unit No 3, the 18-month surveillance testing of the ABFS train "A" fan identified a delay in the start caused by an inherent design characteristic of the flow switch in the circuit. The flow switch is physically located at the train "B" fan to sense train 'B' air flow. Due to the design of the flow switch, the train "A" fan would not be permitted to start inmediately after a LOP. The train "A" ABFS fan started approximately 35 seconds late after receiving a sequenced safeguards signal during the LOP testing. This starting delay was repeated 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> later during the ESF/ LOP test.

• The root cause of this event was a lack of understanding of the flow switch response to a LOP event. Design information provided by the manufacturer of the flew switch was inadequate for engineers to fully understand the unique characteristics of the flow-sensing element. The operations and Instruction manual, a basic design document, did not provide information that would lead a design engineer to understand that flow switch response was susceptible to a subsequent time delay when power was removed from the flow element portion of the switch. Without this information, full understanding of switch operation ,

was inhibited, making it unlikely that the LOP time delay flow would have been caught prior to testing. Additionally, a simulated LOP test performed in October 1992 did not include the flow switch auxiliary circuit. NNEC0 believes this test to be a reasonable one to perform at that time to demonstrate operability.

. w l'c V~ dJ nr.v Jda13 Ve n, f ac t s. 1. l C e D S A D E r nA tw. c 4 000 0000 T.10 DRAFT /U.S. Nuclear Regulatory Commission B14647/Page 14 October XX, 1993 DR "

In the event of an ESF actuation (i.e., upon a SIS) with a LOP, the ABFS also provides an axhaust path to assist the SLCRS in drawing a negative pressure of 0.25 inches water gauge in the secondary containment boundary. Based on previous testing results, the approximate 35 second time delay for starting the ABFS train 'A" fan means that train 'A' of the SLCRS would not draw a negative pressure of 0.25 inches water gauge in the secondary containment boundary within the.60 second requirement. Instead, it is projected that the required negativa pressure would not be reached until approximately 70-80 seconds. At 60 seconds, the secondary containment boundary would have reached a negative pressure between 0.10 and 0.20 inches water gauge. A similar flow switch was functionally deleted from the "B" train of the ABFS in October 1992.

In LER 93-014-00,* NNECO described corrective s.ctions to resolve this condition. A design change was implemented to repower the 3HVR*FNEA flow switch from an uninterruptible power source. Upon completion of the modifications, NNECO performed a ESF/ LOP test on October 11, 1993, to verify system operability. During the performance of this test, the "B" train ABFS fan did not start untti 90 seconds efter a SIS. This failure rendered the SLCRS system inoperable. NNECO's review of this failure has ' identified additional single failura vulnerabilities within the SLCRS instrumentation and controls. To res9lve this issue, a team of enginocrs was assembled to address the cause of the deficiencies in the system. The matter has been pursued on a seven-day per week, extended hour basis as a top corporate priority. NNEC0 has kept the NRC infomed of significant developments in addressing these issues.

Further, the requested emergency authorizaticri is appropriate because this NNECO amendment has doesthat determined not these involveproposed a significant hazards changes areconsideration'(dSHC).

acceptable an thoroughly justified from a safety standpoint.

Sionificant Hazards Consideration In accordance with 10CFR50.92, NNECO has reviewed the attached proposed changes and has concluded that they do not involve an SHe. 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 would not:

1. Involve a sionificant increase _in the orobability or consecuences of an accident orevioush evaluated.

The ability of the combined SLCRS/ABVS to meet the proposed Technical Specification to draw a negative pressure of 0.25 inches water gauge in tie secondary containment boundary within 150 seconds of a start signal (6) S. E. Scace letter to the U.S. Nuclear Regulatory Commission, " Facility Operating " License No. NPF-49, Docket No. 50-423, Licensee Event Report 93-014-00, dated September 30, 1993.

, w rcu u ncu w. m uen racia. ucensing rs Nu, tuo ooo oceo f. le em DRAFT /U.S. Ned ear Regulatory Commission B14647/Page 15 October XX, 1993 DR "

is established through the evaluation of modification-related operating time changes and the use of prior test data. These tests show that the SLCRS/ABVS equipment is capable of developing ull in excess of 0.25 inches water gauge within the secondary containment boundary. There is reasonable assurance that this can be accomplished within the 2.5 minute time period. Furthermore, testing currently will be completed following implementaticrc of previously identified modifications. This testing will validate the system's ability to perfom its intended function in the requisite time frame.

Extension of the time allowed to achieve drawdown of secondary containment from 1 minute to 2.5 minutes will have negligible impact on heating and cooling. Plant experience has shown that heatup and  ;

cooldown of thick-walled concrete structures, such as the Millstone Unit No. 3 auxiliary building, is a relatively slow process. Following an accident signal, ventilation equipment is restarted promptly.

Therefore, heatup or cooldown, during short periods while ventilation fans and/or heaters are inactive, is insignificant and can be neglected.

The sroposed change to decrease the containment integrated leakage rate at t1e design basis pressure from 0.65 wt.%/ day to 0.3 wt.%/ day has been evaluated to determine the impact of the proposed lower leakage criteria on the Millstone Unit No. 3 containment test program. It was determined that the leakage results from the type "A " "B." and 'C' tests for the current refueling outage provide assurance of containment integrity even under the proposed leakage criteria. Also, the results of the bypass leakage are within the proposed limit. The proposed upper bound for the overall integrated leakage of 0.30 wt.%/ day is more restrictive than the current upper bound of 0.65 wt.%/ day.

NNECO has determined that the overall effect of the proposed changes was to reduce the calculated doses. The previously calculated EAB thyroid i and whole body doses were 150 rem and 19.5 rem respectively. . Utilizing the proposed revisions, the EAB doses were calculated to be 146.5 rem thyroid and 9.459 rem whole body. It was also concluded that the total curies of each iodine and noble gas isotope is less over each time period for this analysis than for the current analysis of record. This indicates that the low population zone, control roos, and technical ,'

support center doses will be lower. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the cossib111tv of_ a new or different kind of accident from any accident eraviousiv evaluated.

The proposed changes do not compromise the ability of the SLCRS and ABFS ,

to mitigate the consequences of an accident. Also, the proposed changes do not iavolve any physical alterations to plant equipment or procedures which would introduce any new or unique operaticnal modes or accident r

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DRAFT /U.S. Nuclear Regulatory Commission ,

B14647/Pa0e 16 October XX, 1993 precursors. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a sic,,1ficant reduction in a narain of safety.

NNECO has determined that the overall effect of the proposed changes was to reduce the calculated doses. The previously calculated EAB thyroid and whole body doses were 150 rem and 19.5 rem, respectively. Utilizing the proposed revisions, the EAB doses were calculated to be 146.5 rem thyroid and 9.459 rem whole body. It was also concluded that the total curies of each iodine and noble gas isotope is less over each time period for this analysis than for the current analysis of record. This indicates that the low population zone, control room, and technical support center doses will be lower. Therefore, the proposed changes do not involve a significant reduction in a margin of safety. On the contrary, the proposed changes would slightly increase the margin of safety as gauged by the reduction in the calculated EAB thyroid and whole body doses and the reduction of the total curies of each todine and noble gas isotope for the subject time frames. Further, there is no other parameter affected by this proposed amendment for which it can be concluded that the proposed changes result in a significant reduction in the margin of safety.

Moreover, the Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples (March 6, 1986, 51FR7751) of amendments that are considered not likely to involve as SHC. The proposed change to reduce the acceptance criteria for the overall integrated leac rate required by Technical Specification 3.6.1.2.a is enveloped by example (ii), a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications.

The proposed change to revise Technical Specification 4.6.6.1.d.3 by increasing the time to drew a negative pressura of 0.25 inches water gauge within the secondary containment boundary from 50 seconds to 150 seconds is not enveloped by any of the examples. However, it has been demonstrated that this change concurrent with the change to the upper bound of the overall integrated leakage rate results in a calculated rec luction in the EAB doses to the thyroid and whole body, and a reduction in the low population -zone, control room, and technical support center doses since the total curies of each iodine and noble gas isotope is less over each time period.

Therefore, these proposed changes do not negatively impact the public health or safety, nor do they involve an SHC.

Reouest for an Enforcement Discretton NNECO hereby acknowledges that an alternative available to the NRC in response to this submittal is to exercise discretion not to enforce compliance with the required actions in Millstone Unit No. 3's Tecnnics1 Specifications 3.6.1.2.a,

, w s ' r COO ar.v 40.so ve n. TACab Licensing rnA hv. Gd cod 0030 f. j u DRAFT /U.S. Nuclear Regulatory Commission B14647/Page 17 October XX, 1993 3.6.6.1, and 3.7.9. NNECO is provides justification for enforcement '

discretion associated with the above Technical Specifications.

1. The Technical Seecification Condition that Will Be Violated.

Millstone Unit No. 3 Technical Specification 3.6.6.1 requires the operability of the SLCRS and Technical Specification 3.7.9 requires the operability of the ABFS prior to the plant proceeding to Mode 4.

Technical Specification 3.6.1.2.a is not directly affected, since the recent leakage tests are within the current acceptance criteria, and will be within the proposed limit. The proposed revision to L, is more restrictive than the current limit. Ilowever, the proposed change to Technical Specification 3.6.1.2.a (i.e., reduction in the upper bound of the overall integrated leakage rate) is coupled with the change to Technical Specification 4.6.6.1.d.3.

NNECO is requesting enforcement discretion from the subject Technical Specifications. This discretion should be effective until the amendment is issued and implemented; thus allowing NNECO to start-up and operate Millstone Unit No. 3 in the interlm. ,

2. The circumstances Surroundino the Situation Includino the Need for Premot Action. -

1 As ditcussed in the Background / Sequence of Events Section, NNECO identified the problem with the SLCRS and ABFS during the current (Ccle 4) refueling outage. NNECO notified the NRC in erent design deficiency in the ABF5 in LER 93-014-00. gaff of an In this 4 LER, NNECO describes corrective actions which included design changes to the ABFS. U)on completion of the modifications, NNECO performed a ESF/ LOP test 40 verify system operability. During the performance of the test, the "B" train ABFS fan did not start until 60 seconds after a SIS. This failure rendered the SLCR5 system inoperable. Engineering ,

review of this failure has identified additional single failure vulnerabilities with the SLCRS instrumentation and controls. NNEc0 has been working diligently to reach expeditious resolution of this matter.

3. Safety Basis for the Reauest NNEC0 believes that the safety significance is small and justified. As discussed in the Safety Assessment Section of this letter, the proposed changes do not pose a condition adverse to safety, and there can be no adverse safety consequences created by the proposed changes. The i overall effect of the proposed changes was to reduce the calculated j (7) 5. E. Scace letter to the U.S. Nuclear Regulatory Commission, " Facility Operating License No. NPF-49. Docket No. 50-423, Licensee Event Report 93-014-00," dated September 30, 1993.

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~

DRAFT /U.S. Nuclear Regulatory Comission B14647/Page 18 3 RAFT)i October XX, 1993 doses. The previously calculated EAB thyroid and whole body doses were 150 rem and 19.5 rom, respectively. Utilizing the proposed revisions, the EAB doses were calculated to be 146.5 rem thyroid and 9.459 rem whole body. The evaluation concluded that the total curies of each iodine and noble gas isotope is less over each time period for this analysis than the previous analysis. This indicates that the low population zone, control room, and technical support cer,ter doses will be lower.

4. Cemeensatory Messures The proposed cnforcement discretion would allow NNECO to start-up and operate Millstone Unit No. 3. During the time enforcement discretion applies, the SLCRS and ABFS will be available to perform its safety function. Therefore, no further compensatory actions are deemed necessary.

5. Duration of Recuested Wojver The enforcement discretion is being requested for the period until the license amendment is approved by the NRC. This will allow NNECO to start-up and operate the plant safely.

6. $1 sis for No Sionificant Harneds Consideration The basis for this enforcement discretion not involving an SHC is the same as described previously for the proposed amendment. However, since the period for which enforcement discretion would apply is very brief, the no SHC conclusion is more persuasive.

7. Pasis for No Irreversible Environmental Consecuences The requested enforcement discretion involves no environmental consequences, since the request, if approved, will allow NNECO to start-up and operate Millstone Unit No. 3 safely. The proposed changes result in a reduction in the calculated doses; therefore, they do not negatively impact the public health and safety. The proposed changes do i not affect the associated non radiological effluents, l

8. Safety Review 1he Millstone Unit No. 3 Plant Operations Review Comittee (PORC) and Nuclear Review Board (NRB) have reviewed and approved this request for enforcement discretion.

9. 6Aditional Information Also, Attachment 4 is provided to out11re a third alternative available to the NRC Staff to respond to this request. The third option (the first being issuance of an emergency license amendment, the second being

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1 DRAFT /U.S. Nuclear Regulatory Comission B14647/Page 19

[, i I

October XX, 1993 enforcement discretion as described above) would also involve enforcement discretion but for a rationale unique to the current circumstances and configuration of Millstone Unit No. 3 at this time.

These circumstances are outlined in Attachment 4.

In sumary, the proposed enforcement discretion would allow NNECO to start up and o>erate N111 stone Unit No. 3 safely. This request is safe and does not const1tute an SHC. >

fnvironmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed enanges do not involve an SHC. 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 concludes '

that the proposed changes meet the criteria delineated in 10dFR51.22(c)(9) for a categorical exclusion from the requirements for an environmental impact statement.

The proposed changes to the Technical Specifications presented in Attachments 1 and 2 reflect the currently issued version of the Millstone Unit No. 3 Technical Specifications. ~ The to the Technical Specifications submitted on July 29, proposed 1993.* changes are not reflected in the enclosed retype. NNECO hereby requests the NRC Staff to check for continuity with the Millstone Unit No. 3 Technical Specifications prior to issuance.

The Millstone Unit No. 3 PORC anti NRR have reviewed and approved this proposed '

amendment and concur with the above determination.

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

As discussed in the Introduction and Justification for Emergency License Amendment Sections of this submittal, authorization of these proposed changes is required to support plant start-up. Therefore, NNECO requests that the NRC staff issue the subject amendment prior to October XX,1993, to be effective  !

upon issuance. In parallel, the NRC Staff may wish to consider whether-it is i advisable to exercise enforcement discretion from Technical Specifications a 3.6.6.1, 3.7.9, and 3.6.1.2.a to be effective until the amendment is issued. I The enforcement discretion would permit NNECO to start-up and operate Millstone Unit No. 3 while awaiting approval of the proposed revision to the  ;

Millstone Unit No. 3 Technical Specifications. l (8) J. F. Opeka letter to the U.S. Nuclear Regulatory Comission, ' Millstone Nuclear Power Station, Unit No. 3 Proposed ' Revision to Technical i Specifications, Supplementary Leak Collection and Release System," dated 4 July 19, 1993. ,

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DRAFT /U.S. Nuclear Regulatory Comission N 814647/Page 20 1 *l October XX, 1993 M '

NNECO wishes to emphasize our conclusion that this proposed license amendment does not involve any undue safety risk or irreversible environmental consequences. We are, therefore, requesting this action to allow start-u and eperation of Millstone Unit No. 3. This action is in the interest of the .

health and safety of the public, our customers, and our shareholders.

We will, of course,.promptly provide any additional information the NRC Staff may need to respond to this request. We are also prepared to support a meeting with the NRC Staff at your convenience, if that would be helpful.

Very truly yours, 1 NORTHEAST NUCLEAR ENERGY COMPANY J. F. Opeka i Executive Vice President cct T. T. Martin, Region ! Administrator V. L. Rooney, NRC Project Manager, Millstone Unit No. 3 P. D. Swctland, 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 5065 llartford, Connecticut 06102-5056 Subscribed and sworn to before me  ;

this day of , 1993 Notary Public Date Comission Expires:

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men j

D I - f.

Docket No. 50-423

$14647 t

Attachment 1 Millstone Unit No. 3 Proposed Revision to Technical Specifications Supplementary Leak Collection and Release system Markup Pages ,

October 1993

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C0KTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION l

, 3.6.1.2

• Containment leakage rates shall be limited to:

e J' An overall integrated leakage rate of less than or equal to L , /

0'3 0 psia . psig); a' '

b. A combined leakage rate of less than 0.60 L for all penetrations  :

and valves subject to Type 8 and C tests, when pressurized to P,1 and -

f

c. A combined leakage rate of less than or equal to 0.042 L for all  ;

penetrations identified in Table 3.6-1 es inclosure Bu11d1hg bypass  ;

leakage paths when pressurized to pg .

l 1 APPLItaBitITY: MODES 1, 2, 3, and 4.  !

ACTION:

With the measured overall integrated containeent leakage rate exceeding 0.75 1.lbject to Type B and C tests exceeding 0.60orL.

s theorcombined the seasured bypass combined lea leakage rate exceeding 0.042 L , restore the over81 integrated leakage rate f to less than 0.75 L . the combined leaka e rate for all penetrations subject to Type B and C tesis to less than 0.50 , and the combined bypass leskege .

rate to less than 0.042 L, prior to inc asing the Reactor Coolant system [ t temperature above 200'F. '

KifRVFILiANCF RFnUIRFMENTS e

4 6'.1.2 The containeant leakage rates shall be demonstrated at the following

- test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using methods and provisions of ANSI H45.4-1972 (Total Time Method) and/or ANS!/ANS 56.8 1981 (Mass point Method):

a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40110 month intervals during shutdown at a I pressure not less than P 53.27 psia (38.57 psig) during each i 10 year service period. Ne third test of each set shalls be D conducted during the shutdown for the 10-year plant inservice ,

inspection .

b. If any periodic Type A test fails to meet 0.75 L the test schedule for subiequent Type A tests shall be reviewed In,d approved by the Consission. If two consecutive Type A tests fall to meet 0.75 L , a Type A test shall be performed at least every 18 months until,two  ;

consecutive Type A tests meet 0.75 L" at which time the above test schedule say be resumed: '

i l

1 MILLSTONE - UNIT 3 3/46-2 i

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/ Auguit 22, 1990 CONTAIMMENT SYEIEMS )

EURVE11LANt[ RFoufRIMENTS fcentinued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory i Position C.6.b of Regulatory Guide 1.52 Revision 2, March 1978,* l meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Evide 1.52. Revision 2. March 1978,* for a methyl ,

todido penetration of less than 0.1755;

d. At least since per 18 months by:

Verifying that the pressure drop across the combined HEFA 1) filters and charcoal adsorber banks is less than 6.25 inches [$[

Watercfm 7600 Gauge o 9800while efs, operating the system at a flow rate of g

2) Verifying that the system starts on a Safety In.jection test 3

signal.

3) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus ,

within StriR1md3 after a start signal, and  !

So.setowlo

4) Verifyin]g that the heaters dissipate 50 15 kW when tested in accordance with AN51 N510-1980.

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration /f '

and bypass leakage testing acceptance criteria of less than 0.05% in W/

i accordance with ANSI N510 1980 for a DOP test aerosol while '

operating the systes et a flow rate of 7600 cfm to 9800 cfai and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place ,

penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with AN51 N510-1980 for a halogenated 7 5 /  !

hydrocarbon refrigerant test gas while operating the system at a flow rate of 7600 cfm to g800 cfa. ,

V

1 1

'Ah51 M510 3955 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52 Revision 2. March 1978.

MILLSTONE - UNIT 3 3/4 6-39 Amendment No. g , 53-1,

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. . ' January 31, 1986 CONTAINMENT.5YSTEMS - , l

" ' s BASFS _

3/4.6.6 SECONDARY CONTAINMENT 3/4.6.6.1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEN

]

The OPERABILITV of the supplementary Leak Collection and Release System (sus) ensures that containment leakage occurring during LOCA conditions into the  :

enclosure building will be filtered through the HEPA filters and charcoal j adsorber traine prior to discharge to the atmosphere. Cumulativo operation of--  :

the systes with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of sioisture on the adsorbers and HEPA filters. This requirement is necessary to meet the assusptions used .

in the safety analyses and limit the SITE SOUNDARY radiation doses to within  :

the dose guideline values of 10 CFR Part 100,during LOCA conditions. ANSI ,

N510-1980 will be used as a procedural Guide for surveillance testing.

-. x gg g, 3/4.6.6.2 ENCLOS'JRE RUILDING INTidR1TY _

Secondary CONTAINENT INTEGRITY ensures that the release of radioactive I

. materials from the primary containment atocsphere will be restricted to those l 1eakage paths and associated leak rates assumed in the safety analyses. .

Thisrestriction,inconjunctionwithoperationofthesupplementaryLeak Collection and Release System, will limit the SITE B0UNDARY radiation doses to '

within the dose guideline values of 10 CFR Part 100 during accident conditions.  ;

3/4.6.6.3 ENCLOSURE 8UILDING STRUCTURAL INTEGRITY .!

This ifmitation ensures that the structural integrity of the containment enclosure building will be maintained comparable to the original design stan- l dards for the life of the facility. Structural integrity is requires to pro-vide an annulus surrounding the steel vessel that can be maintained at a  !

negative pressure during accident conditions. A visual inspection is suff t-cient to demonstrate this capability.  ;

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Attachment 2 Millstone Unit No. 3 Proposed Revision to Technical Specifications supplcinentary Leak Collection and Release System Retyped Pages s l l

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October 1993 J

.. w l cu-so wtu w d uen. racil. ucensag rtm NO. 203 660 6es6 P. 2' 8 CONTAINMENT. SYSTEMS CONTAINMENT LEAKAGE { j LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or equal to L.,

0.3% by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at P.,

53.27 psia (39.4 psig):

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b. A combined leakage rate of less than 0.60 L, for all penetrations and valves subject to Type B and C tests, when pressurized to P,;

and

c. A combined leakage rate of less than or equal to 0.042 L. for all penetrations identified in Table 3.6-1 as Enclosure Building bypass leakage paths when pressurized to P.. -

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the measured overall integrated containment leakage rate exceeding 0.75 L,, or the measured combined leakage rate for all penetrations and valves subject to Type 5 and C tests exceeding 0.60 L or the combined bypass leakage rate exceeding 0.042 L , restore the overal,l integrated leakage rate to less than 0.75 L., the combined leakage rate for all penetrations subject '

to Type B and C tests to less than 0.60 L., and the combined bypass leakage rate to less than 0.042 L, prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE REQUIREMENTS 1

4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR Part 50 using methods and provisions of ANSI N45.4-1972 (Total Time Nethod) and/or ANSI /ANS 56.8-1981 (Mass Point Method):

a. Three Type A tests (Overall Integrated Containment Leakage' Rate)  ;

shall be conducted at 40 10 month intervals during shutdown at a pressure not less 10-year service than P},he period. 53.27 thirdpsia test(38.57 psig)set of each during shalleach be conducted during the shutdown for the 10-year plant inservice inspection;

b. If any periodic Type A test fails to meet 0.75 L., the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L., a  !

Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L, at which time the above test schedule may be resumedt .

1 MILLSTONE - UNIT 3 3/4 6-2 Amendment No. JJ ena

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50Rs*EILLANCE REQUIREMENTS (Continued)

c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation, by verifying, within 31 days after removal that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,*

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978,* for a methyl iodide penetration of less than 0.175%:

d. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6.25 inches Water Gauge while operating the system at a flow rate of 7600 cfm to 9800 cfm.

2) Verifying that the system starts on a Safety Injection test signal,

3) Verifying that each system produces a negative pressure of greater than or equal to 0.25 inch Water Gauge in the annulus within 150 seconds after a start signal, and

4) Verifying that the heaters dissipate 50 iS kW when tested in accordance with ANSI N510 1980,

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less than 0.05% in accordance with ANSI N510-1980 for a 00P test aerosol while -

operating the syste:n at a flow rate of 7600 cfm to 98 Cfm; and

f. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing acceptance criteria of less

• than 0.05% in accordance with ANSI N510-1980 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow rate of 7600 cfm to 9800 cfm.

• ANSI N510-1980 shall be used in place of ANSI N510-1975 referenced in Regulatory Guide 1.52, Revision 2, March 1978.

MILLSTONE - UNIT 3 3/4 6-39 Amendment No. F, J7 0173

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3/4.6.6 SECONDARY CONTAINMENT 3/4.6,6.1 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM The OPERABILITY of the Supplementary Leak Collection and Release System (SLCRS) ensures that containment leakage occurring during LOCA conditions into the enclosure building will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. Cumulative operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. This requirement is necessary to meet the assumptions used in the safety analyses and limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during LOCA condittuns. ANSI H510-1980 will be used as a procedural guide for surveillance testing.

The SLCRS is not nomally in operation. The SLCRS starts on a SIS signal. With the SLCHS in postaccident configuration, the required negative pressure in the secondary containment boundary (i.e. the annulus) is achieved in 140 seconds from the time of simulated emergency diesel generator breaker closure. Time delays of dampers and logic delays must be accounted for the 18-month surveillance 4.6.6.1.d.3. The time to achieve the required negative pressure is 150 seconds, with a loss-of-offsite power coincident with a SIS.

3/4.6.6.2 ENCLOSURE BUILDING INTEGRITY Secondary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses.

This restriction, in conjunction with operation of the Supplementary Leak Collection and Release System, will limit the SITE BOUNDARY radiation doses to within the dose guideline values of 10 CFR Part 100 during accident conditions.

3/4.6.6.3 ENCLOSURE BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment enclosure building will be maintained comparable to the original design stan-dards for the life of the facility. Structural integrity is required to. pro-vide an annulus surrounding the steel vessel that can be raintained at a~

negative pressure during accident conditions. A visual inspection is suffi-cient to demonstrate this capability.

MILL 5 TONE - UNIT 3 8 3/4 6-4 Amendment No.

0178

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Attachment 3 Millstone Unit No. 3 Proposed Revision to Technical Specifications Supplementary Leak Collection and Release System Simplified ABVS Drawing 4

October 1993 I

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DRAFT "~"'= 3 Attachment 4 Millstone Unit No. 3 Proposed Revision to Technical Specifications  :

Supplementary Laak Collection and Release System Enforcement Discretion P

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October 1993 1

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• DftAFT/U.S. Nuclear Regulatory Comission i L i j

B / Attachment 4/Page 1 j October XX, 1993  ;

Northeast Nuclear Energy Company (NNECO) hervby requests that the NRC Staff ,

exercise its discretion not to enforce compliance with the required actions in l Millstone Unit No. 3's Technical Specifications 3.6.6.1 and 3.7.9. i Enforcement Discretion is necessary to allow Hillstone Unit No. 3 to proceed to Mode 2 following the current refueling outage to avoid a delay in plant start-up. Permitting the plant to enter Modes 4, 3, and 2 includes establishing normal operating temperature and pressure and enables NNECO to convience surveillances and low power physics tests required upon completion of a refueling outage.

Descrintion of Procesed Enforcement Discretion Technical Specifications 3.6.6.1 and 3.7.9 have applicability in Mode 4.

Therefore, Mode 4 cannot be entered until the Limiting Conditions for Operation (LCO) for both of these Technical Specifications are met. NNECO Is requesting enforcement discretion to allow Millstone Unit No. 3 to proceed up to Mode 2 without having to first demonstrate operability of the supplementary leak collection and auxiliary building filtration system (ABFS). and Therelease maximum system time that ($LCRS)lant the p would be allowed to remain in Mode 2 is seven (7) days before the SLCRS and ABFS would have to be demonstrated c)erable. At that time, the plant would have to proceed to Mode 3 and remain t1ere until the LCOs were met.

Safety Assessment Drawdown of the Millstone Unit No. 3 secondary containment enclosure to a negative pressure of 0.25 inches water gauge is required by Technical Specifications to occur within 1 minute of receiving a safety injection signal. For a design basis loss of coolant accident (LOCA), credit is taken for SLCRS/ABFS performance. One hundred percent of core ncble gases and fifty percent of core iodines are assumed to be instantaneously released to the containment at the start of the accident. Most of the act'vity which leaks into the secondary centainment enclosure is drawn through the ventilation, filtered and reduced by a factor of 20 prior to release. The filters provide no reduction in noble gas activity. Hence, this discussion focuses on the effects on iodine thyroid dose consequences. An assessment of the dose consequences for such an event has concluded that the operation of Millstone Unit No. 3 from Modes 4 through 2 will not produce iodine curies in sufficient quantity to exceed the original filtered dose at 100% power. This assessment assumed that SLCRS and ABFS were unavailable; therefore, the release from the secondary containment enclosure was unfiltered.

Millstone Unit No. 3 has been shutdown for over 80 days for refueling.

Virtually all iodine in the core has decayed away. Only a small fraction (0.001) of the core iodines present at the end of Cycle 4 operation remain.

I During Modes 3 and 4. no additional iodine is produced because the reactor remains subcritical. Since the remaining inventory is significantly less than the 100% power eqailibrium inventory, any release during Modes 3 or 4 even without filtration would be bounded by the 100 % design basis LOCA assuntions.

~

... . .. .. . .. -- ..........a....... .....a.. . . ~ . . ~ .. a M l DRAFT /U.S. Nuclear Regulatory Comission J  ; ,

8 / Attachment 4/Page 2  ;

October XX, 1993 The lodine produced in the core from operating Millstone Unit No. 3 in Mode 2 i (i.e., 5 % power or 1/20th of normal full power) is dependent upon power and time. Operating in Mode 2 for 1 week will result in core DEQ I-131 to be far below 1/20th of the respective curies of DEO I 131 at full power based on the fact that equilibrium at this power will not be achieved for approximately 5 half- lives of the longest lived iodine, I 131 of approximately 8 days. Thus, operating in Mode 2 for 1 week will produce resulting iodines dose consequences for the postulated event below what has already been analyzsd and accepted. The resulting dose consequences for an accident during this operaticn (Mode 2 for 1 week) are bounded by the original design basis analysis.

From analysis of the available curies that could be assumed to be released following a design basis accident LOCA by operating Millstone Unit No. 3 through Modes 2, 3, and 4, it is apparent that the resulting dose consequences would bo less than those calculated for 100% design basis accident LOCA assumptions. Additional insights into the safety significance of this request was provided in the main body of the letter.

DISCUSSION OF IMPACT ON OPERABILITY ON CHARGING AND REACTOR PLANT COMPONENT COOLING WATER Therefore, Millstone Unit No. 3 can be operated in Modes 3 and 4 indefinitely before reaching Mode 2, and upon Hode 2 operation can remain there for 1 week.

This operation is acceptable for one time only following the current (Cycle 4) refueling outage.

Recuest for en Enforcement Discretion NNECO is providing justification for enforcerrent discretion associated with com llance with the Millstone Unit No. 3 Technical Specifications 3.6.6.1, "5u plementary Leak Collection and Release System" and 3.7.9, " Auxiliary But ding Filter System."

1. The Technical Soecification condition that Will Be Violated Millstone Unit No. 3 Technical Specification 3.6.6.1 requires the operability of the SLCRS and Technical Specification 3.7.9 requires the operability of the ABFS prior to the plant proceeding to Mode 4. .

NNECO is requesting enforcement discretion to allow the plant to proceed I UP to Mode 2 to avoid a delay in plant startup.

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DRAFT U.S. Nuclear Regulatory Commission B / Attachment 4/Page3 October XX, 1993

2. The Circumstances Surroundino the Situation Includina the Need for Promot Action This information has been provided in the Background / Sequence of Events Section of the main body of this letter. ,

3. Safety Basis for the Raouest ,

NNECO believes that the safety significance is small and justified. The proposed enforcement discretion would allow Millstone Unit No. 3 to be in Mode 2 for a period not to exceed I week without requiring the SLCRS or ABFS to be operable. Since the unit is starting up from a refueling outage, the inventory of iodines in the core is low, and would not be significantly increased by operation of the core at a maximum of 5%

power for a I week Jeriod. Therefore, the consequences of a design basis accident LOCA curing this period would not exceed those documented in the safety analysis.

STATEMENT IS REQUIRED CONCERNING THE ABILITY TO TAKE ABYS OUT OF SERVICE AND IMPACT ON OPERABILITY OF CHARGING AND REACTOR PLANT COMPONENT COOLING WATER

4. Comeensatory Measures

5. Duration of Reauested hba The enforcement discretion is b1ing requested to allow Millstone Unit No. 3 to be operated in Modes 3 and 4 indefinitely before reaching Mcde ,

2, and upon achieving Mode 2 t<> remain there for the period of I week )

without SLCRS and ABFS operability.  !

6. Bases for No Stanificant Harards Consideration NNEco has reviewed the proposed enforcement discretion in accordance with 10CFR50.92 and has concluded it does not involve a significant hazards consideration. The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not compromised. The proposed waiver l

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DA DRA U.S. Nuclear Regulatory Commission l 5 Attachment 4/Page 4- ,

Octo r XX.-19g3 -

3 j would not Involve a significant hazards consideration because It would -

not: 1

1. Involve a significant Increase In the probability of-occurrence or consequences of an accident previously analyzed. l The proposed enforcement discretion does not change 'the way the unit is" operated. Therefore, it does not affect the probability .

l of any design basis events described -in the Safety Analysis.  !

Report. However, this action would allow entry into Modes 4, 3, and I without two safety systems being operable as required by the. -

current Technical Specifications. These systems directly affect  ;

the off-site dose consequences of design basis events. As ,

discussed above, the unit is returning to service following a  ;

refueling outage. Therefore,- the inventory of todines in the  !

reactor core is very low. This significantly reduces the potential i consequences of any design basis events that would occur during .

startup of the unit.

Also, limiting tho' unit to 55 power for a maximum of one week will not result in a significant buildup of iodines in the core.

Thus, the potential radioloiical consequences of any design basis 1 accidents will be.maintainst below those documented in the Safety j Analysis Report even without the SLCRS and ABFS in operation. 1

2. Create the possibility of a new or different kind of accident from i any previously evaluated. t The possibility of' an accident or malfunction of a different type- ,

than any evaluated previously'in the Safety Analysis Report is not  !

created. --Since there are no changes In -the way the plant is i operated, the potential for an unanalyzed accident is not created. t No new failure modes are introduced.  ;

3. Involve a significant reduction in a margin of safety. l The proposed enforcement discretion does not have any adverse  !

impact on the protective boundaries. The margin of safety, as ';

defined in the basis for any Technical Specification, is not reduced. The reason for this is attributable to the timing and .

restrictions. Since the unit has just completed a 75 day refueling  ;

outage, the core inventory of iodines is low. Also, limiting the- ,

unit to 5% power will not result'in a significant buildup of  :

iodines in the core. The proposed enforcement discretion does affect the performance of two safety' systems. However, the timing ,

and limitations on operation will more than compensate for this j condition. A s

Basis for No Irreversible Environmental conseouances l 7.

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1 ORAFT/U.S. Nuclear Regulatory Comission B . / Attachment 4/Page 5 October XX, 1993 + 3 The proposed enforcement discretion has no environmental impact since the allowance for havirg the SLCRS and ABFS cut of service for the maximum of 7 days while in Mode 2 is more than compensated for by the timing of implementation and the restrictions on plant operation.

in sunnary, the proposed enforcement discretion would permit Millstone Unit No. 3 to enter Mode 2 to avoid a delay in plant startup. This request is safe and does not constitute a significant hazards consideration.

. 8. _ Safety Review PENDING The Millstone Unit No. 3 Plant Operations Review Comittee and Nuclear Review Board have reviewed and approved this request for enforcement discretion.  ;

-