ML20059C771
| ML20059C771 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/27/1993 |
| From: | Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20059C776 | List: |
| References | |
| B14655, NUDOCS 9311020011 | |
| Download: ML20059C771 (24) | |
Text
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HARTFORD CONNECTICUT 06141-0270 k
k J CC.(([l72 (203) 665-5000 October 27, 1993 Docket No. 50-423 B14655 Re:
10CFR50.90 10CFR50.91 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Millstone Nuclear Power Station, C.iit No. 3 Propo. sed Revision to Technical Specifications Supplementary leak Collection and Release System Introduction Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its Operating License, NPF-49, by incorporating the changes identified in Attachment 1 into the Technical Specifications of Millstone Unit No. 3.
The purpose of this submittal is to consolidate the proposed license amendments dated July 29, 1993,"' and October 22, 1993, <2' into a single proposed license amendment to. ensure continuity.
This submittal also incorporates additional information requested by the NRC Staff during ' a conference call conducted on October 19, 1993, and during a meeting conducted on October 25, 1993. These changes principally concern the Millstone Uni _t No.
3 Technical Specifications related to the supplementary leak collection and release system (SLCRS) and auxiliary building filter system (ABFS).
NNEC0 proposes to revise the Millstone Unit No. 3 Technical Specifications as follows:
(1) changing the name of definition section 1.12 from " ENCLOSURE BUILDING INTEGRITY" to " SECONDARY' CONTAINMENT B0UNDARY;"
(1)
J. F. Opeka letter to the U.S. Nuclear Regulatory Commission, " Proposed Revision to Technical Specifications, Supplementary Leak Collection and Release System," dated July 29, 1993.
l (2)
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 October 22, 1993.
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U.S. Nuclear Regulatory Commission B14655/Page 2 Octooer 27, 1993 (2) revising Technical Specifications 3.6.1.2.c, 3.6.6.2, 3.6.6.3, and 4.6.6.3, Table 3.6-1, Bases Sections 3/4.6.6.2 and 3/4.6.6.3, and Index pages (i), (viii), and (ix) to incorporate definition 1.12's new title; (3) eliminating the requirement for SLCRS operability from definition 1.12; (4) 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; (5) delineating in the limiting condition of operation (LCO) for Technical Specification 3.6.6.1 the equipment-necessary to comprise an operable SLCRS; (6) expanding the LC0 for Technical Specification 3.7.9 to denote the equipment necessary to comprise an operable ABFS and to state that the action statements of Technical Specification 3.6.6.1 must be met when an ABFS is declared inoperable; (7) revising Technical Specification 4.6.6.1.d.3 to state that the SLCRS must produce a negative pressure of greater than or equal to 0.40 inches in the auxiliary building at the 24'-6" elevation within 150 seconds after a start signal (this time includes the diesel generator start and load time of approximately 10 seconds);
(8) rewriting Bases Section 3/4.6.6.1 to expand the basis for the SLCRS LCO, action statements, and surveillance requirements, and 4
to clarify the areas represented by the secondary containment boundary; (9) revising Bases Section 3/4.7.9 to.be consistent with the proposed revision to Technical Specification 3.7.9; and (10) replacing the phrase "a halogenated hydrocarbon refrigerant" -in Technical Specifications 4.6.6.1.f, 4.7.7.g and 4.7.9.f with the phrase "an acceptable."
In addition, NNECO is requesting that the NRC Staff process this 'icense amendment request on an emergency basis pursuant to 10CTR50.91(a)(5).
By the current schedule, emergency authorization is required by October 29, 1993, to allow unimpeded operation of the plant.
At the present time, $ 11 stone Unit No. 3 is in the process of ascending in modes following a' refueling outage.
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. operate Millstone Unit No. 3 while the proposed license amendment is being processed.
NNEC0 believes that an emergency iicense amendment is warranted in this case to permit operation of the plart, since the associated operational risk with the request has no negative impact.on public health and safety.
l U.S. Nuclear Regulatory Commission B14655/Page 3 October 27, 1993 At the October 25, 1993, meetin the NRC Staff verbally approved NNECO's request dated October 22, 1993, *g,for enforcement discretion that would allow Millstone Unit No. 3 to be operated in Modes 3 and 4 indefinitely and in Mode 2 for a period not to exceed 7 days without having the SLCRS and ABFS operable.
System Descriotions The operability of the 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 in the t
annulus (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.
The auxiliary building ventilation system (ABVS), which includes the ABFS, provides the normal and postaccident means for cooling 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 filter the exhaust from nonsafety related ventilation systems (e.g.,
containment purge) and during monthly charcoal filter bed surveillance testing.
In the event of a DBA, one train of the ABVS is automatically brought into operation to support the SLCRS and to filter the exhaust of the fans which provide cooling to vital equipment.
The design of the ABFS incorporates two redundant trains of active equipment which share common ductwork and plenums (see Attachment 2 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 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, (3)
Ibid.
i U.S. Nuclear Regulatory Commission B14655/Page 4 October 27, 1993 that the previously operating and now failed (single active failure) 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 independent, even though they are redundant.
Backaround/Seouence of Events 1992 Events On September 29, 1992, the "B" train of the SLCRS was declared inoperable, and e
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 acts in parallel with the SLCRS 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, NNEC0 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 LC0 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, h N,*
NNEC0 described the background, s' 4us, and course of action taken for the resolution of the design l
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 pl ant.
On September 19, 1988, the NRC Staff issued Information Notice No. 88-76 to alert licensees of a phenomenon which could cause the secondary containment
.i pressure to rise above allowable values. Specifically,.the Information Notice i
(4)
J.
F.
Opeka letter to the U.S.
Nuclear Regul atory Commission,
" Supplementary Leak Collection and Release System / Auxiliary Building Ventilation System-Event Summary," dated November 12, 1992.
4 L
U.S. Nuclear Regulatory Commission B14655/Page 5 October 27, 1993 stated that the design of Nine Mile Point Unit No. 2 did not take into account the temperature-induced difference in.the pressure gradients inside and outside the secondary containment.
During periods of time when outside temperatures are significantly lower than reactor building temperatures, pressure differentials between the inside and outside of the secondary enclosure at upper elevations were calculated to be less negative than the allowable value of -0.25 inches water gauge.
Information Notice No. 88-76 stated that the installed instrumentation at Nine Mile Point Unit No. 2 was
" insufficient to accurately determine reactor building differential pressure at higher elevations."
An evaluation determined that Millstone Unit No. 3 was susceptible to the same phenomenon.
The evaluation concluded that the SLCRS was currently operable, and would remain operable through at least October 31, 1993.
This conclusion was based on a comparison of available test data, environmental data presented in the Millstone Unit No. 3 Updated Final Safety Analysis Report, and calculations to account for pressure / temperature differences across the containment enclosure at various elevations.
The evaluation concluded that for various periods of time, particularly during cooler weather, the.SLCRS may have been inoperable.
A prompt report under the auspices of 10CFR50.72(b)(2)(iii)(c) regarding this historical condition was made on June-17, 1993, and a Licensee Event Report (LER 15, 1993.'{' was submitted under the auspices of 10CFR50.73(a)(2)(v)(c) on July 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 duct 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 del ay 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 (ESF)/LOPtest.
L 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 (5)
S. E. Scace letter to the U.S. Nuclear Regulatory Commission, " Facility Operating License No. NPF-49, Docket No. 50-423, Licensee Event Report 93-009-00," dated July 15, 1993.
U.S. Nuclear Regulatory Commission B14655/Page 6 October 27, 1993 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.
A full investigation to determine the impact of the flow switch design on other safety related ventilation systems was undertaken.
In LER 93-014-00,5 NNECO 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.
Task Force 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-h in 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.
The team's review of this failure has identified an additional single failure vulnerability within the SLCRS/ABVS instrumentation and controls and has recommended changes to correct the identified system deficiencies.
Meeting Between NRC Staff and NNECO Conducted on October 25, 1993 On October 25, 1993, NNEC0 representatives met with the NRC Staff at the NRC offices in Rockville, Maryland to provide additional information and assist the NRC Staff in processing our proposed license amendment related to. the SLCRS and the ABFS Technical Saecifications.
NNEC0 presented in detail the i
SLCRS and ABFS requirements anc descriptions, the system testing program and test results, and the failure modes and effects analysis (FMEA) results.
At the end of the meeting, the NRC Staff verbally approved NNEC0's request for enforcement discretion that would allow Millstone Unit No. 3 to be operated in Modes 3 and 4 indefinitely and in Mode 2 for a period not to exceed 7 days without having the SLCRS and ABFS operable.
This submittal is in response to the understandings reached during the ating, in that one comprehensive amendment request is hereby submitted.
Desian Chances The design changes described below have been implemented. They allow specific equipment to start as soon as possible after a LOP event coincident with _ an accident signal.
These modifications will ensure that SLCRS operating in (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," dited September 30, 1993, i
.l U.S. Nuclear Regulatory Commission B14655/Page 7 October 27, 1993 conjunction with the ABVS will achieve drawdown of the secondary containment boundary to a negative pressure-of greater than or equal to 0.40 inches water gauge as measured at the 24'-6" elevation in the auxiliary building within 150 seconds following an accident signal; i.e.,
1.
ABFS Exhaust Fan 3HVR*FN6B Chanaes Under the previous logic, a LOP coincident with a SIS /CDA signal is required in series with high plenum pressure to initiate a 30 second time delay before opening fan 6B's inlet and. outlet dampers.
_These dampers are required to fully open before fan 6B can start. The purpose of this circuit is to detect fan 6A failure and to start the opposite train auxiliary building exhaust fan 6B.
A time delay in this circuit is required to allow fan 6A to start after an accident signal._ However, the time delay (i.e., the window of opportunity 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 a LOP / SIS /CDA signal.
Previous testing has demonstrated when fan 6A starts, the plenum pressure will be reduced prior to the 30-second time period.
In addition, a reset feature was auded 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 68 after an accident condition, two deliberate operator actions must be performed; i.e., reset safety / input signal (LOP, SIS, CDA) and place 3HVR*FN6B control switch in stop.
Under the previous logic, a SIS /CDA signal with both fans (or fan 6B only) running would have restarted fan 6B and stopped fan 6A.
If fan 6B i
had failed, no switchover to fan 6A would have taken place, this would have put the plant in an LC0 condition.
The trip circuit has been revised 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 were running if started manually prior to receiving the safety signal.
2.
Charaina Pumo and Reactor Plant Component Coolina Water (RPCCW) Supp1Y Fan 3HVR*FN14A/B and Exhaust Fan 3HVR*FN13A/B Timina Relav Setooint i
Chanaes This design change revised the timing relay setpoints for fans i
3HVR*FN14A,14B and 13A,13B from 20 seconds to 10 seconds to allow the opposite train fans to start (.arlier on detection of low flow when a fan failure occurs.
This design change reduces 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 6B.
Also, the revised time delay will' i
. -+
U.S. Nuclear Regulatory Commission B14655/Page 8 October 27, 1993 maintain adequate margin for the primary fan to start before the opposite train fan can start.
An operational procedure change to place the fan control switch in the AUTO position has eliminated the potential for operating both trains of charging and RPCCW fans simultaneously due to a flow switch failure or a fan failure.
Operating both trains of 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 Setuoint Chanaes The flow switch setpoint changes allow the flow switches to respond without excessive time delay to a low flow condition caused by a fan failure, yet maintain adequate margin to prevent a spurious low-flow signal when the fans operate normally.
This change reduces the excessive time delay previously experienced for starting fan 6B during the performance of the IST.
System Testina Proaram and Results A series of integrated tests were performed under various normal operating and simulated failure modes, to demonstrate that the ABVS and SLCRS would achieve drawdown to a negative pressure of 0.40 inches water gauge as measured at the 24'-6" elevation in the auxiliary building 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 was lined. up in its various operational modes, with normal power available and the system in service.
System response to the accident signal was then verified for the full spectrum of failures in the power supply system and for certain failures within the system itself.
A number of tests (most limiting) demonstrated the system capability to establish the drawdown condition referenced above.
The sequence of tests permitted the testing to proceed in an orderly manner.
Equipment operation was monitored and the results of each sequence tested were compared with the same evaluated scenario to assure operation of all critical components as expected.
Any emerging issues were resolved before proceeding to other tests.
Although not every scenario evaluated was tested, operation of all conducted scenarios was bounded by scenarios tested, to demonstrate performance of the entire system.
These tests were conducted during the week of October 18, 1993, to demonstrate system capability and operability.
The details of test plans and results were presented at the October 25, 1993 meeting.
N U.S. Nuclear Regulatory Commission B14655/Page 9 October 27, 1993 The test results confirmed that the SLCRS/ABVS will achieve drawdown to a negative pressure greater than or equal to 0.40 inches water gauge as measured at the 24'-6" elevation in the auxiliary building within 150 seconds following an accident signal.
NNECO is confident that the pertinent set of tests comprising the integrated test program are all-encompassing and that the-
- system, when subjected to a comprehensive test series, has responded appropriately.
This was a probing test series which established a detailed understanding of how the combined systems operate and respond to any credible accident scenario.
The results of the Winter Mode test will be communicated to the NRC Staff as soon as they are available.
Description of Proposed Chanael NNEC0 proposes to revise the Millstone Unit No. 3 Technical Specifications as follows:
(1) changing the name of definition section 1.12 from " ENCLOSURE BUILDING INTEGRITY" to " SECONDARY CONTAINMENT BOUNDARY;"
(2) revising Technical Specifications 3.6.1.2.c, 3.6.6.2, 3.6.6.3, and 4.6.6.3, Table 3.6-1, Bases Sections 3/4.6.6.2 and 3/4.6.6.3, and Index pages (i), (viii) and (ix) to incorporate definition 1.12's new title; t
(3) eliminating the requirement for SLCRS operability from definition i
1.12; (4) 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; (5) delineating in the LC0 for Technical Specification 3.6.6.1 the equipment necessary to comprise an operable SLCRS; (6) expanding the LCO for Technical Specification 3.7.9 to denote the equipment necessary to comprise an operable ABFS and to state that the action statements of Technical Specification 3.6.6.1 must be met when an ABFS is declared inoperable; l
(7) revising Technical Specification 4.6.6.1.d.3 to state that the SLCRS must produce a negative pressure of greater than or equal to 0.40 inches in the auxiliary building at the 24'-6" elevation within 150 seconds after a start signal (this time includes the diesel generator start and load time of approximately 10 seconds);
(8) rewriting Bases Section 3/4.6.6.1 to expand the basis for the SLCRS LCO, action statements, and surveillance requirements, and i
to clarify the areas - represented by the secondary containment boundary; (9) revising Bases Section 3/4.7.9 to be consistent with the proposed revision to Technical Specification 3.7.9; and (10) replacing the phrase "a halogenated hydrocarbon refrigerant" in Technical Specifications 4.6.6.1.f, 4.7.7.g and 4.7.9.f with the phrase "an acceptable."
U.S. Nuclear Regulatory Commission B14655/Page 10 October 27, 1993 NNECO proposes to define the phrase " secondary centainment boundary" in the definitions section of the Millstone Unit No. 3 Technical Specifications.
Specifically, definition 1.12, " ENCLOSURE BUILDING INTEGRITY" will be renamed
" SECONDARY CONTAINMENT B0UNDARY."
Renaming definition 1.12 requires that Technical Specifications 3.6.1.2.c, 3.6.6.2, 3.6.6.3, 4.6.6.3, Table 3.6-1 and Bases Sections 3/4.6.6.2 and 3/4.6.6.3 be reworded to achieve consistency throughout the Technical Specifications.
Also, Index pages (i), (viii), and (ix) will need to be revised to reflect the changes to Technical Specifications 3.6.6.2 and 3.6.6.3, and Table 3.6-1.
These changes are editorial in nature.
NNEC0 proposes to delete the SLCRS operability requirements from definition 1.12 because this requirement is already covered in Technical Specification 3.6.6.1, " Supplementary Leak Collection and Release System."
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 Technical Specifications has been proposed to enable NNECO to also revise Technical Specification 4.6.6.1.d.3 as discussed below.
NNEC0 proposes to revise the LC0 for Technical Specification 3.6.6.1 to state that an operable SLCRS is comprised of one operable filter and fan, and one operable ABFS.
The revision to the LC0 for this Technical Specification acknowledges that the c:CRS works in conjunction with the ABFS to achieve a negative pressure within the secondary containment boundary.
This proposed change would impose an additional restriction on the operation of Millstone Unit No. 3.
NNECO proposes to revise the LC0 for Technical Specification 3.7.9 to delineate that an operable ABFS is comprised of one operable filter and fan, and one operational charging pump / reactor plant component cooling water pump ventilation system.
The proposed change to Bases Section 3/4.7.9 reflects this change. Also, NNEC0 proposes to modify the action statement of Technical Specification 3.7.9 by denoting that the action requirements of Technical Specification 3.6.6.1 must be met when an ABFS is declared inoperable.
The proposed changes to Technical Specification 3.7.9 would impose additional restrictions on the operation of Millstone Unit No. 3.
NNEC0 proposes to revise Technical Specification 4.6.6.1.d.3 by changing the surveillance acceptance criterion from a negative pressure greater than or equal to 0.25 inches water gauge in the annulut within 50 seconds after a start signal (this time does not include the diesel generator start. and load time of approximately 10 seconds) to a negative pressure greater than or equal to 0.40 inches water gauge in the auxiliary building at the 24'-6" elevation within 150 seconds after a start signal.
The proposed time requirement includes the diesel generator start and load time of approximately 10 seconds.
This proposed acceptance criterion will ensure a negative pressure in all
U.S. Nuclear Regulatory Commission B14655/Page 11 October 27, 1993 areas within the secondary containment boundary under most meteorological conditions.
The acceptance criterion is more restrictive due to the consideration of temperature effects noted in Information Notice 88-76.
Measuring the pressure at the 24'-6" elevation of the auxiliary building is adequate and representative of the entire secondary containment boundary due to the large cross-section of the air passage which interconnects the various buildings within the boundary.
To support these changes, NNEC0 proposes to replace Bases Section 3/4.6.6.1 with a discussion that would be more informative.
The guidance of NUREG-1431* was utilized to develop the proposed rewrite of Bases Section 3/4.6.6.1.
Also, the proposed rewrite of Bases Section 3/4.6.6.1 will define the areas comprising the secondary containment boundary, as well as provide an expanded basis for the LCO, action statements, and surveillance requirements.
These changes will add to the information available to individuals concerned with the operability of the SLCRS.
NNECO proposes to replace the phrase "a halogenated hydrocarbon refrigerant" denoted in Technical Specification Surveillances 4.6.6.1.f, 4.7.7.g, and 4.7.9.f with the phrase "an acceptable. "
This proposal will provide Millstone Unit No. 3 with additional flexibility concerning which gas may be used to leak rate test the charcoal filters.
In the near future, the use of certain halogenated hydrocarbons as test gases will be banned, lherefore, the industry will have to adopt a new standard test gas.
This change simply prepares Millstone Unit No. 3 for this future event.
No change to the acceptance criterion is being proposed; thus, the assumed efficiency of the charcoal filters will be unaffected.
Safety Assessment Appendix A to Section 15.6.5 of the Standard Review Plan (SRP)'S and Regulatory Guide 1.4* provide the guidelines for performing the calculation of radiological consequences given a loss of coolant accident (LOCA).
These documents establish a set of simplifying nonmechanistic assumptions to use in the calculation.
Utilizing these assumptions permits one to reach the J
conclusion specified in the SRP that the distances to the exclusion aroa boundary (EAB) and to the low population zone (LPZ) of the site, in conjunction with the ESF of the plant are sufficient to provide reasonable assurance that the total radiological consequences of such an accident will be (7)
NUREG-1431, " Standard Technical Specifications, Westinghouse Plants,"
issued by the U.S. Nuclear Regulatory Commission in September 1992.
(8)
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition," issued by the U.S.
Nuclear Regulatory Commission in June 1987.
l (9)
" Assumptions Used for Evaluating the Potential Radiological Consequences of a LOCA for PWRs," Revision 2, June 1974.
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U.S. Nuclear Regulatory Commissior, B14655/Page 12 October 27, 1993 within the exposure guidelines set forth in 10CFR Part 100, 100,11.- The key
)
words in this conclusion are " reasonable assurance."
To assume credit for the secondary containment filtration systems, the_SRP states that the secondary containment must reac'n a negative pressure differential of 0.25 inch water gauge.
The SRP does not provide a basis for the magnitude of the negative pressure, nor does it delineate any specific requirements regarding where the pressure differential must be measured.
As discussed in tha Background Section of this submittal, Information Notice No. 88-76 informed licensees of a previously unidentified phenomenon which could cause the secondary containment pressure to rise above allowable values.
Basically, the Information Notice states that if a licensee is measuring the pressure differential at lower elevations, then the potential exists for the licensee to be at an unacceptable pressure at higher elevations in a tall building during periods of time when the outside and inside air temperatures are significantly different.
The Information Notice did not state where pressure differential measurements should be taken, or calculational guidance to assume a given percent exfiltration if specific negative pressure differentials could not be achieved.
A detailed review of the Millstone Unit No. 3 design drawings has shown that the contiguous buildings defined to be within the secondary containment boundary, namely the ESF building, the auxiliary building (AB), the hydrogen recombiner building (HRB), and the main steam valve (MSV) building, are interconnected via open air spaces (called shake spaces) to the enclosure building.
These connections have a large total cross section, as shown in the following table.
Each of the buildings, except for the MSV building, are further connected to the SLCRS. suction by ductwork.
Building HRB AB MSV ESF Conn. Area, sq. ft.
24 46 38 56 i
SLCRS Duct. Conn.
Yes Yes No Yes l
Top Elev.
47'-10" 91'-6" 84'-11" 54'-9" An evaluation determined that the various sections of the secondary containment boundary can be considered as one large space due to the large interconnections and ductwork.
Any pressure measurement in one location will represent the pressure in other locations at the same elevation.
Since the top of the enclosure building (approx, el.186') is higher than the top elevation of the other buildings, the calculation for establishing the j
acceptance criterion only dealt with the enclosure building.
Indoor j
i
a >
abA.:a 4
4 i
U.S. Nuclear Regulatory Commission B14655/Page 13 October 27, 1993 temperature differences between buildings will have no effect on the acceptance criterion, because the limiting case is the enclosure building.
With the SLCRS in a postaccident configuration, the negative pressuro in the secondary containment boundary 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 in this surveillance.
Thus, the time to achieve the required negative pressure is 150 seconds, with a LOP coincident with a SIS.
The proposed changes to Technical Specification 4.6.6.1.d.3 will verify that one train of SLCRS in conjunction with the ABFS will produce a negative pressure of 0.4 inches water gauge relative to the outside atmosphere at the 24'-6" elevation of the auxiliary building within 150 seconds of a start signal.
For the purpose of this surveillance, pressure measurements will be made at the 24'-6" elevation in the auxiliary building.
Measuring the negative pressure at this single location is adequate and representative of the entire secondary containment boundary due to the large cross-section of the air passages which interconnect the various buildings within the boundary.
In nrder to ensure a negative pressure in all areas inside the secondary containment boundary under most meteorological conditions, the acceptance criterion at the measured location is a negative pressure of 0.4 inches water gauge.
It is recognized that there will be an occasional meteorological condition during which slightly positive pressure may exist at some localized portions of the boundary (e.g., the upper elevations on the down wind side of a building).
For example, a very low outside temperature combined with a moderate wind speed could cause a slightly positive pressure at the upper elevations of the containment enclosure building on the leeward face.
Based upon historical
- data, the probability of occurrence of meteorological conditions which could result in such a positive differential pressure condition in the upper levels of the enclosure building has been estimated to be less than 2% of the time.
The probability of wind speed within the necessary moderate band, combined with the probability of extreme low temperature, the small portion of the boundary affected, and the low probability of airborne radioactive material migrating to the upper levels ensure that the overall affect on the design basis dose calculations is insignificant.
The ability of the SLCRS and ABVS to meet the proposed Technical Specification requirement to draw a negative pressure of 0.40 inches water gauge at the 24'-6" elevation of the auxiliary building within 150 seconds of the start of an accident has been demonstrated, as described in the meeting conducted on October 25, 1993.
To further assure operability, a FMEA of the ABVS was conducted to confirm the absence of 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*FN6A/68, m
U.S. Nuclear Regulatory Commission B14655/Page 14 October 27, 1993 3HVR*FN13A/138, and 3HVR*FN14A/14B and the flow switches 3HVR*FS278, 3HVR*FS52A/52B, and 3HVR*FS98A/98B 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 mal 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 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,,
j when L, would be equal to 0.30 wt.%/ day).
1
U.S. Nuclear Regulatory Commission B14655/Page 15 October 27, 1993 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 for the EAB and the LPZ as depicted in the table below.
LOCATION / DOSE TYPE FSAR NEW EAB 0-2 HR Thyroid 150 REM 146.5 REM EAB 0-2 HR Whole Body 19.5 REM 9.5 REM LPZ 30-DAY Thyroid 31.6 REM 18.1 REM LPZ 30-DAY Whole Body 3.5 REM 1.7 REM In addition, 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 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.
Justification for Emeraency License Amendment Pursuant to 10CFR50.91(a)(5), NNEC0 hereby requests NRC Staff " emergency" approval of the proposed amendment to its Operating License NPF-49.
Under the current schedule, emergency authorization is required by October 29, 1993, to allow entry into Mode 1.
At the present time, Millstone Unit No. 3 is in the process of ascending in modes following a refueling outage.
i 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.
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 duct 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
U.S. Nuclear Regulatory Commission B14655/Page 16 October 27, 1993 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 be 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 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 "A" ductwork of the ABFS in October 1992.
In LER 93-014-00,"* NNECO 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, NNECO performed a SLCRS drawdown 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 the system.
The matter has been pursued seven days per week, on an extended-hour basis as a top corporate priority. NNEC0 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).
NNEC0 j
(10)
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 B14655/Page 17 October 27, 1993 has determined that these proposed changes are acceptable and thoroughly justified from a safety standpoint.
Sianificant Hazards Consideration 1
In accordance with 10CFR50.92, NNECO 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 consecuences of an accident previously evaluated.
The ability of the SLCRS and ABVS to meet the proposed Technical Specification to draw a negative pressure of 0.40 inches water gauge in the auxiliary building at the 24'-6" elevation within 150 seconds after a start signal (this time includes the diesel generator start and load time of approximately 10 seconds) is established through the evaluation of modification-related operating time changes and the use of prior test data.
The test results confirmed that the SLCRS/ABVS will achieve drawdown to a negative pressure of 0.40 inches water gauge as measured at the 24'-6" elevation in the auxiliary building within 150 seconds
+
following an accident signal. NNECO is confident that the pertinent set of tests comprising the integrated test program are all-encompassing and that the system, when subjected to a comprehensive test series, has rosponded appropriately.
This was a probing test series which established a detailed understanding of how the combined systems operate and respond to any credible accident scenario.
These tests show that the SLCRS and ABVS equipment are capable of developing a negative pressure in excess of 0.40 inches ' water gauge within the auxiliary building at the 24'-6" elevation.
There is reasonable assurance that this can be accomplished within 150 seconds (this time includes the diesel generator start and locd times).
Furthermore, testing recently completed following implementation of previously identified modifications validated the system's ability to j
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 j
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.
4
-r-w
i
~
U.S. Nuclear Regulatory Commission B14655/Page 18 October 27, 1993 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 j
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 th{nin2roid 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 LPZ, 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 an accident previously evaluated.
The proposed change to delineate the equipment required to comprise an operable SLCRS in the LCO for Technical Specification 3.6.6.1 would impose an additional restriction on the operation of Millstone Unit No. 3.
The proposed change to delineate the equipment necessary to comprise an operable ABFS in the LC0 for Technical Specification 3.7.9, and the proposed change to require compliance with the action statements of Technical Specification 3.6.6.1 when an ABFS is declared inoperable in the action statements for Technical Specification 3.7.9 will impose additional restrictions on the operation of Millstone Unit No. 3.
The proposed revision to Bases Section 3/4.7.9 will ensure that the basis and the LCO for Technical Specification 3.7.9 are consistent with each -
i other.
i The proposal to replace the phrase "a
halogenated hydrocarbon refrigerant" denoted in Technical Specification Surveillances 4.6.6.1.f, 4.7.7.g, and 4.7.9.f with the phrase "an acceptable" will not change any of the acceptance criteria for the charcoal filter leak rate test.
The j
(11)
E.
J.
Mroczka letter to the U.S.
Nuclear Regulatory Commission,
" Proposed Revision to Technical Specifications, Containment Pressure,"
dated February 26, 1990.
(12)
D. H. Jaffe letter to E. J. Mroczka, " Issuance of Amendment No. 59 (TAC No. 76066)," dated January 25, 1991.
i
~-
U.S. Nuclear Regulatory Commission B14655/Page 19 October 27, 1993 changes are being proposed to permit Millstone Unit No. 3 to cope with the future ban of certain halogenated hydrocarbons.
Because the changes will not impact the assumed efficiency of the charcoal filters, the calculated dose consequences of any postulated accident will not be affected.
The proposal to rename definition section 1.12 as " SECONDARY CONTAINMENT BOUNDARY,"
and the resultant changes to Technical Specifications 3.6.1.2.c, 3.6.6.2, 3.6.6.3, and 4.6.6.3, Table 3.6-1, and Bases Sections 3/4.6.6.2, and 3/4.6.6.3 are editorial in nature.
They do not have any safety impact.
Also, removal of the SLCRS operability requirement from definition 1.12 is acceptable since this requirement is addressed by Technical Specification 3.6.6.1.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes do not compromise the ability of the SLCRS and ABFS to mitigate the consequences of an accident.
A FMEA confirmed that the design changes implemented do not introduce any new single failure vulnerabilities.
The proposed changes do not introduce any new or unique operational modes or accident 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 sianificant reduction in a marain 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 LPZ, 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 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, i
51FR7751) of amendments that are considered not likely to involve a significant hazards consideration.
Several of the proposed changes are i
enveloped by example (ii), a change that constitutes an additional limitation, i
U.S. Nuclear Regulatory Commission B14655/Page 20 October 27, 1993 restriction, or control not presently included in the technical specifications. They are: (1) defining the equipment necessary to comprise an operable SLCRS in the LCO of Technical Specification 3.6.6.1; (2) delineating the equipment necessary to comprise an operable ABFS in the LC0 of Technical Specification 3.7.9; (3) requiring compliance with the action requirements of Technical Specification 3.6.6.1 when the action statements of Technical Specification 3.7.9 are entered; and (4) reducing the acceptance criterion for the overall integrated leak rate required by Technical Specification 3.6.1.2.a.
The proposed rewrite of Bases Section 3/4.6.6.1 is not enveloped by any of the specific examples.
This change will provide a more informative discussion of the basis for the LCO, action statements, and surveillance requirements.
The proposed changes to Technical Specification Surveillances 4.6.6.1.f, 4.7.7.g, and 4.7.9 f to replace the phrase "a
halogenated hydrocarbon refrigerant" with "an acceptable" is similar to example (vii); a change to conform a license to changes in the regulations where the license change results in very minor changes to facility operations clearly in keeping with the regulations. The proposal to rename definition Section 1.12, and the resultant changes to Technical Specifications 3.6.1.2.c, 3.6.6.2, 3.6.6.3, and 4.6.6.3, Table 3.6-1 and Bases Sections 3/4.6.6.2 and 3/4.6.6.3 are enveloped i
by example (i), a purely administrative change to technical specifications; i
for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature.
The proposed change to revise Technical Specification 4.6.6.1.d.3 by increasing the time to draw a negative pressure of 0.40 inches water gauge within the auxiliary building at the 24'-6" elevation from 60 seconds to 150 seconds (these times include the diesel generator start and load time of approximately 10 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 reduction in the EAB doses to the thyroid and whole body, and a reduction in the LPZ, control room, and technical support center doses since the total curies of each iodine and noble gas 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.
1 Reauest for Enforcement Discretion NNEC0 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 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.
9 U.S. Nuclear Regulatory Commission B14655/Page 21 October 27, 1993 l.
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 the overall integrated leakage rate) is coupled with the change to Technical Specification 4.6.6.1.d.3.
At the October 25, 1993, meeting, the NRC Staff verbally approved NNEC0's request dated October 22, 1993, for enforcement discretion that would allow Millstone Unit No. 3 to be operated in Modes 3 and 4 indefinitely and in Mode 2 for a period not to exceed 7 days without having the SLCRS and ABFS operable.
NNECO is requesting enforcement discretion from the subject Technical Specifications to allow Millstone Unit No. 3 to enter Mode 1.
This discretion is requested to be effective until the amendment is issued and implemented, thus allowing NNECO to operate Millstone Unit No. 3 in the interim.
2.
The Circumstances Surroundina the Situation Includina the Need for Prompt Action As discussed in the Background / Sequence of Events Section, NNECO identified the problem with the SLCRS 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, NNECO describes corrective actions which included design changes to the ABFS.
Upon completion of the modifications, NNEC0 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 Reauest NNEC0 believes that the safety significance is small and justified.
As discussed in the Safety Assessment Section of this letter, the proposed (13)
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 B14655/Page 22 October 27, 1993 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 revisions, the EAB doses were calculated to be 146.5 rem thyroid and 9.459 rem whole body.
The evaluation concluded that the i
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 LPZ, control room, and technical support center doses will be lower.
4.
Compensatory Measures The proposed enforcement discretion would allow NNEC0 to operate Millstone Unit No. 3 in Mode 1.
During the time enforcement discretion applies, the SLCRS and ABFS will be available to perform its safety l
function.
Therefore, no further compensatory actions are deemed necessary.
5.
Duration of Reauested Waiver The enforcement discretion is being requested for the period until the license amendment is issued by the NRC.
This will allow NNEC0 enter Mode 1 and operate the plant safely.
6.
Basis for No Significant 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 Conseauences The requested enforcement discretion involves no environmental consequences, since the request, if approved, will allow NNECO to 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 not affect
{
the associated non-radiological effluents.
8.
Safety Review i
The Millstone Unit No. 3 Plant Operations Review Committee (PORC) and i
Nuclear Review Board (NRB) have reviewed and approved this request for enforcement discretion.
l U.S. Nuclear Regulatory Commission B14655/Page 23 October 27, 1993 9.
Additional Information Additional information has been supplied throughout the text of this submittal.
In summary, the proposed enforcement discretion would allow NNECO to enter Mode 1 and operate Millstone Unit No. 3 safely. This request is safe and does not constitute a SHC.
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 statement.
The Millstone Unit No. 3 PORC and NRB 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
.I Amendment Sections of this submittal, authorization of these proposed changes j
is required to support entry into Mode 1 and subsequent plant operation.
.i Therefore, NNECO requests that the NRC Staff issue the subject amendment on or before October 29, 1993, to be effective upon issuance. As an alternative for l
an interim period, the NRC Staff may wish to consider exercising enforcement discretion from Technical Specifications 3.6.1.2.a, 3.6.6.1, and 3.7.9 to be ef fective until the amendment is issued.
The enforcement discretion would permit NNECO to enter Mode 1 and operate Millstone Unit No. 3 while' awaiting issuance of the proposed revision to the Technical Specifications.
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 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.
i
=
i i
U.S. Nuclear Regulatory Commission B14655/Page 24 October 27, 1993 We-will, of course, promptly provide any additional information the NRC Staff.
may need to respond to this request, and we appreciate your ongoing efforts in support of this request.
l Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY
't FOR:
J. F. Opeka Executive Vice President BY:
M Od3]A E. A. DeBarba Vice President cc:
T. T. Martin, Region I Administrator V. L. Rooney, NRC Project Manager, Millstone Unit No. 3 P. D. Swetland, Senior Resident Inspector, Millstone Unit Nos.1, 2, and 3 Mr. Kevin T.A. McCarthy, Director Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street P.O. Box 5066 Hartford, Connecticut 06102-5066 i
f Subscribed and sworn to before me this d day of (kIMw,1993 k ccl. Y O u Notap Public i
Date Commission Expires: a/3//93 i
.-.