ML20059C294
| ML20059C294 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/22/1993 |
| From: | Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20059C299 | List: |
| References | |
| B14647, NUDOCS 9311010084 | |
| Download: ML20059C294 (21) | |
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October 22, 1993 Docket No. 50-423 B14647 Re:
10CFR50.90 10CFR50.91 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 d
Gentlemen:
Millstone Nuclear Power Station, Unit No. 3 Preposed Revision to Technical Specifications Supolementary leak Collection 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 (this. time does B91 include the diesel generator start and load time of approximately 10 seconds) to 150 seconds (this time will include the -diesel generator start and ' load time of~
approximately 10 seconds), and reduce the allowable integrated containment leakage rate (L.) 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 the design basis pressure, respectively. Also, NNECO is proposing to revise Bases Section 3/4.6.6.1 by adding an 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 10CFR50.91(a)(5).
Emergency authorization is. required by October 25, 1993, to support start-up (i.e., entry into Mode 4) i of-the plant. At the present time, Millstone Unit No. 3 is shutdown (Mode 5).
In parallel with this effort, the NRC Staff may wish to consider whether it is-advisable to exercise enforcement discretion-from Technical 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.
NNEC0 believes that an emergency license amendment is warranted in this case to permit the start-up and operation of the plant, since the associated 010007 (0
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U.S. Nuclear Regulatory Commission B14647/Page 2 October 22, 1993 operational risk with the request has no negative impact on public health and safety.
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 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.
System 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 release to the environment.
The SLCRS system is a two-train filtration system with a common inlet and discharge 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 auxiliary 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 proposed changes to the Millstone Unit No. 3 Technical Specifications related to the SLCRS and auxiliary building filtration system (ABFS).
The principal purpose of the proposed changes is to address the phenomenon described in Information 28 Notice 88-76 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 obtained; delineate the equipment required to comprise an operable SLCRS; and denote the equipment required to comprise an operable ABFS.
Also, NNEC0 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 (1)
J. F. Opeka letter to the U.S. Nuclear Regulatory Commission, " Millstone Nuciear 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 Operating Licenses or Construction Permits for Nuclear Power Reactors, "NPC Information Notice No. 88-76:
Recent Discovery of a Phenomenon Not Previously Considered in the Design of Secondary Containment Pressure Control," dated September 19, 1988.
(
U.S. Nuclear Regulatory Commission B14647/Page 3 October 22, 1993 Section 3/4.6.6.1. 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 at certain times during the year.
In that i
submittal, NNEC0 requested that the NRC Staff = issue the subject amendment prior to October 31, 1993.
We hereby reaffirm 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 within 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 mode 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 simplified diagrams).
Each train of the ABVS is powered via redundant and independent power supplies.
There are no normal or accident modes of operation wherein it is acceptable to automatically and simultaneously operate both trains of the ABVS.
- However, both ABFS fans do operate at the same time in conjunction with some nonsafety-related auxiliary building fans.
This further complicates the ABVS/ABFS control schemes.
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 negative 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/Seouence 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 the ABVS fan start from signal actuation.
The ABVS system acts in parallel with the SLCRS and can affect the
e U.S. Nuclear Regulatory Commission B14647/Page 4 October 22, 1993 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 SLCRS drawdown surveillance was large enough to consider the surveillance inadequate for verifying system operability.
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 on 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 IST results showed that the 0.25-inches negative pressure criterion could not be met in the required 60 seconds [80 seconds actual]).
NNEC0 began a shutdown of the unit.
The
-shutdown to Mode 5 was completed on October 1,1992.
In a letter dated November 12,1992,* NNEC0 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 prior to the start-up of the plant.
Current Refueling Outage During the current (Cycle 4) refueling outage at Millstone 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 immediately 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 containment boundary would have reached a negative pressure between 0.10 and 0.20 inches (3)
J.
F.
Opeka letter to the U.S.
Nuclear Regulatory Commission,
" Supplementary Leak Collection and Release System / Auxiliary Building Ventilation System-Event Summary," dated November 12, 1992.
~
U.S. Nuclear Regulatory Commission B14647/Page 5 October 22, 1993 i
water gauge.
A similar flow switch was functionally deleted from the "B"
train of the ABFS 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 Report (LER) 93-014-00,* NNEC0 described corrective actions to resolve this condition.
On October 11, 1993, NNEC0 performed a SLCRS drawdown test to verify system operability upon completion of the modification.
During the 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.-
i Task Force NNECO's review of this failure has identified additional single failure vulnerabilities within the SLCRS/ABVS 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 seven days per week, on - an extended-hour basis as a top corporate priority.
The team consists of representatives from the Probabilistic Risk Assessment Section, the Project Services Department, the Engineering Department, Plant Engineering, and Nuclear Licensing.
Desian Chanaes The design changes described below are being implemented and will allow specific equipment to start as soon as possible 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 150 seconds following an accident i
signal; i.e., LOP, SIS, containment depressurization actuation signal (CDA).
1.
ABFS Exhaust Fan 3HVR*FN6B Chanaes i
Under the present logic, a LOP coincident with ~ the SIS /ESF signal is required in series with a high plenum pressure to initiata 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 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 i
pressure switch in the plenum.
However, the time delay (i.e.,
the window of cpportunity for fan 6A to start) has been changed to begin at the time of the event.
The logic change starts this 30-second time delay at the initiation of the LOP / SIS /ESF signal.
Previous testing has i
(4)
S. E. Scace letter to the U.S. Nuclear Regulatory Commission, " Facility i
Operating License No. NPF-49, Docket No. 50-423, Licensee Event Report 93-014-00," dated September 30, 1993.
t U.S. Nuclear Regulatory Commission B14647/Page 6 October 22, 1993 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 6B control switch 1A-3HVR*FN6B located at the main ventilation panel (3HVS*PNLVPI). This feature will ensure that in order to shut down fan 6B after an accident condition, two deliberate operator actions must be performed; i.e., reset safety signal LOP, SIS, CDA and place 3HVR*FN6B control switch in stop.
Under the present logic, a SIS /CDA signal with both fans (or fan 6B only) running will restart fan 6B and stop fan 6A.
If fan 6B fails, no switchover to fan 6A will take place, which will put the plant in an LC0 condition.
The proposed design will revise the trip circuit by replacing timer 62B - 3HVR*FN6B with an auxiliary relay.
This change will allow the shutdown of fan 6B and the start of fan 6A first, if both fans (or fan 6B only) were running prior to receiving the safety signal.
2.
Charaina Pumo and Reactor Plant Component Coolina Water (RPCCW) Sucolv f_qn 3HVR*FN14A/B and Exhaust Fan 3HVR*FN13A/B Timina Relay Setooint Chanan 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 a quicker start of fan 68. Also, the revised time delay will maintain adequate margin for the primary fan to start before the opposite train fan can start.
The operation procedure change for the fan control switch in AUTO position will eliminate 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.
Charaina Pumo and RPCCW Area Supolv Fans 3HVR*FN14A/B and Exhaust Fans 3HVR*FN13A/B Flow Switch Setooint Chanaes The flow switch setpoint changes will allow the flow switches to respond without excessive time delay to the low flow condition caused by a fan failure yet maintain adequate margin to prevent a spurious low flow signal when the fan operates normally.
This change will reduce the excessive time delay experienced for starting fan 6B during the IST.
U.S. Nuclear Regulatory Commission j
B14647/Page 7 October 22, 1993 l
System Testina Proaram and Results A series of integrated tests will be performed under various normal and failure modes, to demonstrate that the ABVS and SLRCS will achieve drawdown of the secondary containment boundary to a negative 0.25 inches water gauge within 150 seconds following an accident signal (this time includes the diesel generator start and load time of approximately 10 seconds).
For each test performed, the system will be lined up in its various operational modes, with normal power available and the system in service.
System response to the accident signal will then be verified for the full spectrum of failures in the power supply system and for certain failures within the system itself.
A required number of tests (most limiting) will also demonstrate the system capability to establish the drawdown condition referenced above.
The sequence of tests will permit the testing to proceed in an orderly manner.
Equipment operation will be monitored and the results of each sequence tested will be compared with the same evaluated scenario to assure operation of all critical components is as expected.
Any emerging issues will be resolved before proceeding to other tests.
Although not every scenario evaluated will be tested, operation of all conducted scenarios will be bounded by scenarios tested to demonstrate performance of the entire system.
Further, these tests are being conducted during the week of October 18, 1993, to demonstrate system capability and operability.
The details of test plans and results will be presented at the October 25, 1993 meeting.
Descriotion of Proposed Chances NNEC0 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 will include the diesel generator start and load time of approximately 10 seconds), and by revising Bases Section 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 0.25 inches water gauge within the secondary containment boundary.
NNEC0 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 0.65 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.
This proposed change is more restrictive than the current Technical Specification requirement.
This revision to the
j s
l U.S. Nuclear Regulatory Commission B14647/Page 8 October 22, 1993 1
Technical Specifications has been proposed to enable NNECO to also revise Technical Specification 4.6.6.1.d.3.
NNEC0 proposes to revise Technical Specification 4.6.6.1.d.3 by-increasing the time required to draw a negative pressure of 0.25 inches water gauge within the secondary containment boundary from 50 seconds (this time does nat include the diesel generator start and load time of approximately 60 seconds) to 150 seconds (this time includes the diesel generator start and load time of approximately 10 seconds).
In addition, NNEC0 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 Technical Specifications presented in Attachments 1 and 2 reflect the currently issued version of the Millstone Unit No. 3 Technical Specifications.
The proposed changes to the Technical Specifications submitted on July 29, 1993,* 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, NNEC0 requested that the NRC Staff issue the subject amendment prior to October 31, 1993.
In the near future, NNEC0 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 SLCRS and 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 the start of an accident is established through the evaluation of modification-related operating time changes and the use of prior test data.
These tests show that the SLCRS and ABVS equipment is capable of developing in excess of 0.25 inches water gauge within the secondary containment boundary. There is reasonable assurance that this can be accomplished within 150 seconds (this time includes the diesel generator start and load time of approximately 10 seconds).
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 modifications will demonstrate that the (5)
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.
1 U.S. Nuclear Regulatory Commission B14647/Page 9 October 22, 1993 4
systems are operable, and that any potential single failure vulnerabilities j
have been satisfactorily addressed.
i To further assure operability, a failure modes and effects analysis (FMEA) of the ABVS using the probabilistic safety assessment (PSA) process was conducted to determine if there are any single failure vulnerabilities. ABVS components whose operation could potentially affect the SLCRS performance were included in this analysis.
Special emphasis was placed on those components which brought about interactions between the two trains of ABVS.
The ABVS fans 3HVR*6A/B, 3HVR*13A/B, and 3HVR*14A/B the flow switches 3HVR*FS278, 3HVR*FS52A/B, and 3HVR*FS98A/B are some of the significant components that were included in the FMEA.
Since SLCRS is a standby system that is needed subsequent to a release into the secondary containment boundary, the analysis focused on SIS and LOP initiators.
Due to the complexity of the system, it was necessary to clearly identify a list of 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.
No other credible single failures were identified.
Extension of the time allowed to achieve drawdown of secondary containment from 60 seconds to 150 seconds (these times include the diesel generator start and load time of approximately 10 seconds) 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.
Also, natural convection within the auxiliary building tends to stabilize temperatures.
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 leakage results were 0.096 wt.%/ day and 0.084 wt.%/ day, respectively.
These are significantly below the current Technical Specification requirement of 0.39 wt.%/ day (0.6 L.,
when L, is equal to 0.65 wt.%/ day), and would be below the proposed limit of 0.18 wt.%/ day (0.6 L.,
when L, would be equal to 0.30 wt.%/ day). Also, the results for bypass leakage were 0.007 wt.%/ day for as-found and 0.008 wt.%/ 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 1
U.S. Nuclear Regulatory Commission B14647/Page 10 October 22, 1993 Specification limit of 0.0126 wt.%/ day (0.042 L, when L, would be equal to 0.30 wt.%/ day).
In addition, the results from the type "A"
test were 0.13268 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 0.225 wt.%/ day (0.75 L,
when L, would be equal to 0.30 wt.%/ day).
NNEC0 has evaluated the above changes to determine the impact they would have on off-site doses during a design basis loss of coolant accident (LOCA).
The overall effect of the proposed changes was to reduce the calculated doses.
The previously calculated exclusion area boundary"(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. The evaluation concluded that the total curies of each iodine and noble gas isotope is less 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, NNECO hereby provides the information contained in this section. We believe that this information is important because it puts into perspective the risk significance of the SLCRS failure to meet its current design requirement within the described time interval of 60 seconds (this time includes the diesel generator start and load time of approximately 10 seconds).
Given the conservatisms in the design and analysis, the potential safety significance of not meeting the 60 second time interval is essentially zero because potential releases would be minimized as follows.
Please note that our proposed revision to Technical Specifications contains an additional conservatism that is not considered in the following discussion (the proposed change to the upper bound for the overall integrated leakage rate):
Using conservative DBA assumptions, a delay in the drawdown 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 (6)
E.
J.
Mroczka letter to the U.S.
Nuclear Regulatory Commission,
" Proposed Revision to Technical Specifications, Containment Pressure,"
i dated February 26, 1990.
(7)
D. H. Jaffe letter to E. J. Hroczka, " Issuance of Amendment No. 59 (TAC No. 76066)," dated January 25, 1991.
U.S. Nuclear Regulatory Commission B14647/Page 11 October 22, 1993 release that would remain below 10CFR100 dose limits; i.e.,
using Regulatory Guide'l.109 dose factors, the thyroid dose is 286 rem.
If the same conditions are assumed, but ICRP 30 dose factors are utilized, then the thyroid dose would be 180 rem.
Design basis LOCA radiological evaluations are performed based upon 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 resulted 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.
Unfortunately, this simplifying non-mechanistic assumption was applied to the mechanistic design requirements of the radioactivity reduction systems.
This has resulted in some unnecessarily restrictive design requirements for these systems.
One example, applicable to this Technical Specification, is the requirement that in order to assume zero filter bypass, the secondary containment boundary must be drawn to a negative pressure of 0.25 inches water gauge within 60 seconds of. the LOCA.
The 60-seconds requirement only exists because of the non-mechanistic assumption that core melt levels of radioactivity are instantaneously leaking from the containment.
Another non-mechanistic, unrealistic assumption is that until a negative pressure of 0.25 inches water gauge is achieved, all activity leaking into the secondary containment 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 injection 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 into Containment Under Design Basis and Severe Accident Conditions,"
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 Power Plants" 1992 Draft, provide a more probable time for significant fission product release for a pressurized water reactor of 30 minutes.
A negative pressure of 0.25 inches water gauge could be established well within this time period.
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
U.S. Nuclear Regulatory Commission B14647/Page 12 October 22, 1993 released in the DBA calculations.
Hence, as long as the secondary containment boundary is sufficiently negative by the time we expect any significant radioactivity leakage into the enclosure, there should be no 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 released unfiltered.
It is expected that under most conditions, the secondary containment boundary will be at a negative pressure of 0.25 inches water gauge within 60 seconds.
For ~ the potential scenarios where this may 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-minute time frame.
Hence, there is no reason to believe any measurable fraction of the activity leaking into the secondary containment boundary will be transported to the outside walls and leak through any small openings.
Probabilistic Risk Assessment Insights Regarding the SLCRS and ABFS 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 Probabilistic Safety Study (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 ruptureduetoovergressure).
The consequences for such accidents are in the range of 10+ to 10+ man-rem.
The 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 failure occurred, the most likely reason is that it was because containment heat removal systems (sprays) had also failed.
In turn, this is most likely caused by loss of support systems such as all AC power.
Therefore, SLCRS/ABVS are also likely failed for the risk dominant accident sequences.
The SLCRS and ABVS 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 l
containment sprays. Much of that dose is from noble gases.
U.S. Nuclear Regulatory Commission B14647/Page 13 October 22, 1993 The 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 design basis 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 person-rem assumes no credit for-secondary filtration.
Hence, this would be the maximum potential reduction in risk if all containment leakage (vice containment failure) was eliminated.
SLCRS will not eliminate the noble gases which contribute most of the 2 person-rem.
In addition, the first five 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 risk of the change is zero.
We request that the NRC consider the above perspectives when responding to NNEC0's request in this submittal. NNECO is' fully committed to a conservative operating philosophy, and believes that the above considerations provide substantial evidence to that effect.
Justification for Emeraency License Amendment Pursuant to 10CFR50.91(a)(5), NNECO hereby requests NRC Staff " emergency" approval of the proposed amendment to its Operating License NPF-49.
Emergency authorization is required by October 25, 1993, to allow start-up (entry into Mode 4). At the present time, Millstone 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 intricacies of system design and their interrelationships have contributed significantly to the difficulties NNEC0 has encountered in dealing with the issues discussed in this submittal.
U.S. Nuclear Regulatory Commission B14647/Page 14 October 22, 1993 During the current (Cycle 4) refueling outage at -Millstone 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 immediately 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 flow 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 flaw 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 believed this test to b9 a reasonable one to perform at that time to demonstrate operability.
In the event of an ESF actuation (i.e., upon a 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 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 negative pressure would not be reached until approximately 70-80 seconds. At 60 seconds, the semrjary 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,* NNEC0 described corrective actions to resolve this condition.
A design change was implemented to repower the 3HVR*FN6A flow switch from an uninterruptible power source.
Upon completion of the modifications, NNEC0 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 until 90 seconds after a SIS.
An unacceptable accumulation of time delays rendered the SLCRS. system inoperable. To resolve this issue, a team of engineers was assembled to address the cause of the deficiencies in (8)
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.
U.S. Nuclear Regulatory Commission B14647/Page 15 October 22, 1993 the system.
The matter has been pursued seven days per week, on an extended-hour basis as a top corporate priority.
NNECO has kept the NRC informed of significant developments in addressing these issues.
Further, the requested emergency authorization is appropriate because this amendment does not involve a significant hazards consideration (SHC).
NNECO has determined that these proposed changes are acceptable and thoroughly justified from a safety standpoint.
Sionificant Hazards Consideration In accordance with 10CFR50.92, NNEC0 has reviewed the attached proposed changes and has concluded that they do not involve an 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 would not:
1.
Involve a sianificant increase in the orobability or conseauences of an accident oreviously evaluated.
The ability of the SLCRS and ABVS to meet the proposed Technical Specification 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 and ABVS equipment is capable of developing a negative pressure well in excess of 0.25 inches water gauge within the secondary containment boundary.
There is reasonable assurance that this can be accomplished within 150 seconds (this time includes the diesel generator start and load times).
Furthermore, testing currently underway will be completed following implementation of previously identified modifications.
This testing will validate the system's ability to perform its intended function in the requisite time frame.
Extension of the time allowed to achieve drawdown of secondary containment from 60 seconds to 150 seconds (these times include the diesel generator start and load time of approximately 10 seconds) 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.
Also, natural convection within the auxiliary building tends to stabilize temperatures.
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 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
I U.S. Nuclear Regulatory Commission B14647/Page 16 October 22, 1993 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.
NNEC0 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 I
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 increase in the probability or consequences of l
an accident previously evaluated.
4 1
2.
Create the possibility of a new or different kind of accident from any j
accident oreviously 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 involve any physical alterations to plant equipment or procedures l
which would introduce any new or unique operational modes or accident precursors.
Therefore, the proposed changes do not create the i
possibility of a new or different kind of accident from any accident j
previously evaluated.
3.
Involve a sianificant reduction in a maroin of safety.
NNEC0 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 1
not involve a significant reduction in a margin of safety.
On the contrary, the proposed changes would slightly increase the margin of j
1 (9)
E.
J.
Mroczka letter to the U.S.
Nuclear Regulatory Commission,
" Proposed Revision to Technical Specifications, Containment Pressure,"
dated February 26, 1990.
(10)
D. H. Jaffe letter to E. J. Mroczka, " Issuance of Amendment No. 59 (TAC No. 76066)," dated January 25, 1991.
_.___-------a-
U.S. Nuclear Regulatory Commission B14647/Page 17 j
October 22, 1993 safety as gauged by the reduction in the calculated EAB thyroid -and whole body doses and the reduction of the total curies of each iodine 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 an SHC. The proposed change to reduce the acceptance criteria for the overall integrated leak 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 draw a negative pressure of 0.25 inches water gauge within the secondary containment boundary from 60 seconds to 150 seconds (these times include the diesel generator start and load time of approximately 10 seconds) is not envelop?d 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 lea'(age rate results in a calculated reduction 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 pas isotope is less over each time period analyzed.
Therefore, these proposed changes do not negatively impact the public health or safety, nor do they involve an SHC.
Reauest for an Enforcement Discretion 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 Hillstone Unit No. 3's Technical Specifications 3.6.1.2.a, 3.6.6.1, and 3.7.9.
NNECO hereby provides justification for enforcement discretion associated with the above Technical Specifications. Of course, the validity of this request is contingent upon successful completion of the design modifications and testing program described herein.
1.
The Technical Specification 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.
However, the proposed change to Technical Specification 3.6.1.2.a (i.e., reduction in the upper bound of
U.S. Nuclear Regulatory Commission B14647/Page 18 October 22, 1993 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 interim.
2.
The Circumstances Surroundina the Situation Includina the Need fpr Prompt Action As discussed in the uackground/ Sequence of Events Section, NNECO identified the problem with the SLCR3 and ABFS during the current (Cycle
- 4) refueling outage.
NNECO notified the NRC Staff of an inherent design deficiency in the ABFS in LER 93-014-00.""
In this
-LER, NNEC0 describes corrective actions which included design changes to the ABFS.
Upon completion of the modifications, NNECO performed a ESF/ LOP test to verify system operability.. During the performance of the test, the "B" train ABFS fan did not start until 90 seconds after a SIS.
This failure rendered the SLCRS system inoperable.
Engineering review of this failure has identified additional single failure-vulnerabilities with the SLCRS/ABVS instrumentation and controls. NNEC0 has been working diligently to reach expeditious resolution of this matter.
3.
Safety Basis for the Reouest NNECO 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 overall effect of the proposed changes was to reduce the calculated doses.
In addition, the previously calculated EAB thyroid and.whole body doses were 150 rem and 19.5 rem, respectively.
Utilizing the proposed revisiors, 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 center doses will be lower.
(11)
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.
~
U.S. Nuclear Regulatory Commission B14647/Page 19 October 22, 1993 4.
Compensatory Measures The proposed enforcement 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 Reauested Waiver i
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 saf ely.
6.
Basis for No Sianificant Hazards 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.
Basis for No Irreversible Environmental Consecuences The requested enfarcement 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 1
not affect the associated non-radiological effluents.
j 8.
Safety Review The Millstone Unit No. 3 Plant Operations Review Committee (PORC) and Nuclear Review Board (NRB) have reviewed and approved this request for enforcement discretion.
9.
Additional Information i
Also, Attachment 4 is provided to outline 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 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 summary, the proposed enforcement discretion would allow NNECO to start-up and operate Millstone Unit No. 3 safely.
This request is safe and does not constitute an SHC.
-.. -=
~-. -. - -.
~ ~.
i
~
U.S. Nuclear Regulatory Commission 4
B14647/Page 20 October 22, 1993 Environmental Considerations NNECO has reviewed the proposed license amendment against the-criteria of 10CFR51.22 for environmental considerations.
The proposed changes 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 10CFR51.22(c)(9) for a categorical exclusion from the requirements for an environmental impact 1
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 proposed changes to the Technical Specifications submitted on July 29, 1993,"*'
are not reflected in the enclosed retype.
NNEC0 hereby requests the NRC Staff to check for continuity a
with the Millstone Unit No. 3 Technical Specifications prior to issuance.
The Millstone Unit No. 3 PORC and NRB have reviewed and approved this proposed.
1 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.
t t
As discussed in the Introduction and Justification for Emergency License Amendment Sections of this submittal, authorization of these proposed changes-
'1 is required to support plant start-up. Therefore, NNEC0 requests that the NRC Staff issue the subject amendment prior to October 25, 1993, to be effective upon issuance.
In parallel, the NRC Staff may wish to consider whether it'is advisable to exercise ' enforcement discretion from Technical Specifications.
3.6.1.2.a, 3.6.6.1, and 3.7.9 to be effective until the amendment is issued.
The enforcement discretion would permit NNECO to start-up. and operate Millstone Unit No. 3 while awaiting approval of the proposed revision to the Technical Specifications.
3 i
NNEC0 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-up and operation of Millstone Unit No. 3.
This action is in the interest of the health and safety of the public, our customers, and our shareholders.
1 j
(12)
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.
I
00T-22-93 FR1 14:0'2 NUCLEERi10ENSING FAX NO. 2036655376 P.23 ~
U.S. Nuclear Regulatory Commission B14647/Page 21 October 22, 1993 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, NORTHEAST NUCLEAR ENERGY COMPANY r A1 J. U peka
/
Executive Vice (P' resident cc:
T. T. Hartin, Region I Administrator V. L. Rooney, NRC Project Manager, Millstone Unit No. 3 P. D. Swetland, Senior Resident Inspector, Hillstone Unit Nos.1, 2, and 3 Mr. Kevin T.A. McCarthy, Director Honitoring and Radiation Division Department of Environmental Protection 79 Elm Street P.O. Box 5066 Hartford, Connecticut 06102-5066 Subscribed and sworn to before me r tiiiT W - day of 6e 7 4, 1993 cL k h s
Notary Public
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Date Commission Expires: _
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