ML20210A372
| ML20210A372 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/28/1987 |
| From: | Mroczka E NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20210A376 | List: |
| References | |
| B12269, NUDOCS 8705050074 | |
| Download: ML20210A372 (6) | |
Text
'
' s, V
'{'~ "
~
}
e O
O General Offices
- Seiden Street, Berlin, Connec'icut EsEEYsYaNcow P.O. BOX 270
- s ** *** co*"
HARTFORD CONNECTICUT 06141-0270 L
L J *,*,"]$ [,M,'co#",
(203) 665-5000 April 28,1987 Docket No. 50-245 B12269 Re: 10CFR50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:
Millstone Nuclear Power Station, Unit No.1 Proposed Revision to Technical Specifications Primary Containment Isolation Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its Operating License, DPR-21, by incorporating the changes identified in Attachment 1 into the Technical Specifications of Millstone Unit No.1.
The proposed change will revise Technical Specification Table 3.7.1, " Primary Containment Isolation," which identifies the primary containment isolation valves. The change is being made to reflect plant modifications (i.e., valves which are no longer primary containment boundaries, addition of new systems, valve relocation, etc.) and to increase the scope of the table to include not only valves which receive a containment isolation signal, but also check valves and valves opened during power operation for testing and/or sampling purposes.
Technical Specification Table 3.7.1 was originally intended to be a listing of all ralves which receive a containment isolation signal or shut automatically during a containment design basis accident (DBA). This table was never a complete list of all Millstone Unit No. I containment valves, it includes valves which are no longer containment boundaries, but still receive containment isolation signals.
These points are clarified by the change to the Technical Specification Bases, Section 4.7.D presented in Attachment 1.
Many of the valves being added to Table 3.7.1 and two valves which are being deleted are associated with the following new or modified systems: Post Accident Sampling System (PASS); Containment Hydrogen / Oxygen Analyzer System; Drywell Nitrogen Compressor System; Scram Discharge Instrument Volume Modifications; and Control Rod Drive System Modifications. Since the safety concerns of these modifications were already addressed prior to their implementation, the following discussion will only address the safety implications of the valve positions and operation.
M e7eeoso m e,o,e DR ADOCK 05000245 PDR
U5 Nucl:ar Regulatory Commission B12269/Page 2 April 28,1987 Table 3.7.1 will be presented in a new format, containing a section listing automatic isolation valves and another listing manual isolation valves.
AUTOMATIC ISOLATION VALVES Drywell Nitrogen Compressor System - The primary nitrogen suction isolation valve (AC-40) and its backup valve (AC-41) were added with the new drywell nitrogen compressor system. They both close on a Group II isolation. They are both located outside of containment. These two new valves are being added to the table, along with the nitrogen purge stop valve (AC-17), which has always been present. AC-17 is the second containment valve in the lines leading to the purging vaporizer, it is normally closed and it falls closed.
Containment Hydrogen / Oxygen Analyzer System - Eight new valves used in this system are being added to the table. They are the torus sample inboard and outboard isolation valves (AC-194 and AC-195), the lower drywell sample inboard and outboard isolation valves (AC-196 and AC-197), the upper drywell sample inboard and outboard isolation valves (AC-198 and AC-199), and the effluent return to drywell outboard and inboard valves (AC-205 and A C-206),
respectively. These valves are permitted to be open or closed during plant operations. The open or closed option is listed to permit arbitrary alignment of one of three sample paths. Once a sample path has been selected, it is lef t on line until instrument calibration is required. All of these valves go to their closed positions when a < ontainn ent (Group II) isolation signal is transmitted.
Scram Discharge Instrument Volumes - Modifications to the scram discharge instrument volume piping have nevessitated the addition of eight valves to the table. The vent isolation valves (SIG-IN and SDV-IS) and their backups (SDV-2N and SDV-25) and the drain isolati>n valves (SDV-3N and SDV-35) and their backups (SDV-4N and SDV-4S) isolate the scram discharge instrument volumes from their vent and drain paths. On a scram, the scram vent valves open and permit these instrument volumes and the scram discharge headers to become extensions of containment. The previously mentioned eight valves which are being added to Table 3.7.1 close on a scram and thus become the containment isolation boundary under these conditions. It should be noted that a Group 11 isolation signal does not directly close these valves. However, a Group 11 signal willinitiate a scram, which will cause the valves to close.
Recirculation Sample Line - Two recirculation loop sample line valves (SM-1 and SM-2) have been replaced and renamed (RR-36 and RR-37). The inner valve was moved to a position outside of the drywell. In NUREG-0324 Section 4.20.4, the NRC concluded that there is no significant change in failure probability at a penetration having both containment valves outside of containment.
TIP System - Four TIP ball valves and the TIP purge check valve are being added to Table 3.7.1.
The ball valves close automatically (af ter probe retraction is completed) on a Group 11 isolation. These valves are backed up by the TIP squib valves which can be fired to ensure line closure if a cable retraction or ball valve closure failure occurs. The purge check valve is the only containment boundary in its containment penetration. This configuration is acceptable (per Regulatory Guide 1.11) because the purge line can be classified as an instrument line with a built-in flow restriction (due to its smallinside diameter).
1 l
i
O U.S. Nuclear Regulatory Commission B12269/Page 3 April 28,1987 Reactor Water Cleanup System - The bypass valve (CU-2A) around the cleanup system inlet valve (CU-2) is being added to the table. It is normally closed and stays closed on a Group II isolation.
Feedwater System - The normal position for the feedwater check valves (FW-9A, FW-10A, FW-9B and FW-10B) is being changed from open to "N/A".
This is because the check valves are positioned by process flow, not by a normal plant logic feature or operational lineup.
Standby Liquid Control System - Two check valves are provided in series to isolate the standby liquid control system from the reactor. One check valve (SL-
- 8) is located inside the drywell and the other (SL-7) is outside the drywell. The normal position for both of these valves is being changed from closed to "N/A" because their position is determined by system process flow.
Atmospheric Control System - The vacuum relief air to secondary containment valves (AC-2A and AC-28) are being added to Table 3.7.1.
These large check valves have always been present as backup containment isolation valves for AC-3A and AC-3B. They ensure containment isolation capability if AC-3A or AC-3B fails open.
Reactor Head Cooling System - The normal position of the inner isolation valve (HS-5) is being changed from closed to "N/A" because the valve position is determined by system process flow.
REMOTE MANUAL ISOLATION VALVES Post Accident Sampling System (PASS) - Four valves within the PASS system, the Shutdown Cooling System isolation valve and its backup (PAS-17 and PAS-20) and the torus isolation valve and its backup (PAS-24 and PAS-25), are being added to Table 3.7.1. These valves are normally closed. They must be opened periodically during plant operation to permit reactor coolant samples to be drawn. This operational sampling is necessary to maintain the qualifications of chemistry personnel required to use the PASS. These valves do not close automatically on a Group II isolation. This does not create the potential for an increase in the off-site consequences of a DBA because of the following reasons:
These valves and all of the tubing up to the sample cabinet are either o
Category I or capable of being designated as Category I (i.e., certified materials test reports are available). All lines and valve supports have been designed to withstand a seismic event. Therefore, these lines will not break open during the simultaneous occurrence of a DBA and an earthquake.
o The reactor coolant sample cabinet has not been seismically analyzed.
Consequently, the integrity of lines in this cabinet af ter a DBA and an earthquake cannot be assumed, even though engineering judgment leads to the conclusion that this cabinet would endure these events without line breakage. Reactor coolant samples are cooled to 1650F before they reach this cabinet. A leak at the cabinet would result in a spill without off-site dose consequences.
Back leakage from the reactor coolant system would be prevented by check valves in the sample return lines.
All sample supply and return lines are isolable from the control room or
U,1 Nucl:ar R;gulatory Commission B12269/Page 4 April 28,1987 the RCS remote valve operating panel (by means of additional Category I remotely actuated valves).
Core Spray, Low Pressure Coolant Injection (LPCI), and Atmospheric Control Systems - The Core Spray admission valves (CS-5A and CS-3B), the LPCI inboard stop valves to loop injection (LP-10A and LP-10B) and torus spray (LP-14A and LP-14B), the LPCI outboard stop valves to drywell spray (LP-15A and LP-ISB),
and the torus to reactor building vacuum breaker valves (AC-3A and AC-3B) are being added to Table 3.7.1.
These valves are cycled quarterly to demonstrate their functionability per the Millstone Unit No.1 inservice inspection (ISI) program and Technical Specification Surveillance Requirement 4.13.
All of these valves a e normally closed during operation and open to recover from a DBA. The L1-CI and core spray valves are in high integrity piping loops that qualify as closed loops off of containment. Automatic valve positioning logic networks will reposition these LPCI and core spray valves to accomplish recovery from a DBA (i.e., automatic features will override the manual open position switches).
Valves AC-3A and AC-3B are automatically opened to prevent excess depressurization of the containment during a DBA. Having these valves manually open during a DBA will not interfere with their vacuum relief function. Leakage out of the torus through AC-3A and AC-3B will be stopped by containment check valves AC-2A and AC-2B until operator action can place AC-3A and AC-3B back in the automatic mode.
Control Rod Drive (CRD) System - The CRD hydraulic return line has been rerouted and returns to the "A" feedwater line. In this new position, it is outside of the feedwater containment boundary. The old CRD hydraulic return line was cut and capped outside the primary containment upstream of the isolation check valves, during a refueling outage in 1978. During the 1980 refueling outage, the reactor vessel nozzle and the drywell penetration (X-36) for the old CRD return piping were capped and the remaining piping inside the drywell was removed.
Since they have been removed, and are no longer in service, the control rod hydraulic return check valves (301-95 and 301-98) will be deleted from Table 3.7.1.
MANUAL ISOLATION VALVES This section is being added to Table 3.7.1 in the form of a note to permit the repositioning of manual isolation valves for the purpose of special tests or surveillances under the direct control of administratively-approved procedures.
The BWR Standard Technical Specifications contain a similar section.
NNECO has reviewed the attached proposed changes pursuant to 10CFR50.59 and has determined that they do not constitute an unreviewed safety question.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety (i.e., safety-related) previously evaluated in the Final Safety Analysis Report have not been increased. The possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report has not been created.
There has not been a reduction in the margin of safety as defined in the basis for any Technical Specification. These proposed changes will not result in physical changes to the plant or changes in the way the plant is operated. It adds valves to a table which identifies primary containment isolation valves. The two valves which will be deleted from the table are both check valves that are no longer in service.
l U.S. Nucirr Regulat::ry Commission B12269/Page 5 April 28,1987 NNECO has reviewed the proposed changes, in accordance with 10CFR50.92, and has concluded that they do not involve a significant hazards consideration in that these changes would not:
1.
Involve a significant increase in the probability or consequences of an accident previously analyzed. There are no physical changes to the plant as a result of the proposed changes, therefore, previously - analyzed accidents are not affected.
2.
Create the possibility of a new or different kind of accident from any previously analyzed. There are no changes in the way the plant is operated, therefore the potential for an unanalyzed accident is not created.
3.
Involve a significant reduction in a margin of safety. Adding additional valves to Table 3.7.1 imposes more restrictive surveillance requirements for primary containment isolation valves. Since this change imposes an ad:litional surveillance requirement, it does not reduce the margin of safety as specified in the basis of any Technical Specification. The two valves being deleted from Table 3.7.1, CRD valves 301-95 and 301-98, are no longer in service.
The Commission has provided guidance concerning the application of standards in l
10CFR50.92 by providing certain examples (51 FR 7750, March 6,1986). The changes proposed herein are enveloped by example (ii), a change that constitutes an additional control not presently included in the Technical Specifications, in that the addition of valves to Table 3.7.1 constitutes an additional control as the valves will fall under the survei. lance requirement for all primary containment isolation valves. The changes involving the deletion of CRD valves 301-95 nd 301-98 from Table 3.7.1 as discussed above, most closely resemble example (i), a purely administrative change to Technical Specifications, justified by the fact that these valves are no longer in service.
The Millstone Unit No.1 Nuclear Review Board has reviewed and approved the attached proposed revision and has concurred with the above determinations.
In accordance with 10CFR50.91(b), we are providing the State of Connecticut with a copy of this proposed amendment.
Pursuant to the requirements of 10CFR170.12(c), enclosed with this amendment request is the app!! cation ice of $150.00.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY E. 3.4-4! M W czka ~
' ice Preside [nt/
Seni
tid. Nucl:ar Rngulatory Commission B12269/Page 6 April 28,1987
- 3. M'. Allan, Acting Region I Administrator cc:
- 3. 3. Shea, NRC Project Manager, Millstone Unit No.1
. T. Rebelowski, Resident Inspector, Millstone Unit No. I Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, Connecticut 06116 STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD
)
Then personally appeared before me E. 3. Mroczka, who being duly sworn, did state that he is Senior Vice Presidt nt of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensee herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
AAA/d htte Notary Pu My Commission Expires March 31,1988
<