ML20199L283
| ML20199L283 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 04/08/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8604100338 | |
| Download: ML20199L283 (9) | |
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Public Service Electric and Gas Ccmpany Ctrbin A. McNeill, Jr.
Public Service Electric and Gas Company P.O. Box 236, H ancocks Bridge. NJ 08038 609 339-4800 Wce Presrdent -
Nuclear April 8, 1986 Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention:
Ms. Elinor Adensam, Director Project Directorate 3 Division of BWR Licensing
Dear Ms. Adensam:
HIGH PRESSURE-LOW PRESSURE INTERFACE HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Pursuant to a telecon between representatives of Public Service Electric and Gas Company (PSE&G) and the NRC on April 3, 1986 concerning the impact of a fire on high pressure-low pressure interfaces, PSE&G confirms the following:
1.
With one (1) safety relief valve stuck open the core will not be uncovered.
2.
The valves in the one inch (l") bypass lines around testable check valves in the Residual Heat Removal System (RHR) and Core Spray System (CS) shall have power removed during normal operation.
3.
The RHR suction valves (BC-HV-F008 and F009) identified in Hope Creek FSAR Section 9A response to 10CFR50, Appendix R, Section'III.G.2 criteria states that the subject valves have power removed'during normal operation.
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.i Director of Nuclear 2
4-8-86 Reactor Regulation i
A complete review of the impact of a-fire on all high pressure-low pressure interfaces has been performed.
Modifications i
required to the FSAR based on this review of the aforementioned items are attached.
In the event there are any additional concerns in this regard do not hesitate to contact us.
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l Sincerely, 4
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i Attachments i
C D.H. Wagner j
USNRC Licensing Project Manager R.W.
Borchardt
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USNRC Senior Resident Inspector I
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Drywell Pressure (HC Category 1; RG Category 1)
Position:
Implemented.
See A6, B9, C8, C10, and D4.
BB.
Drywell Sump Level (HC Category 3; RG Category I's Position:
Implemented as Category 3.
See C6 and Issue 4, Section 1.8.1.97.4.4.
d.
Maintaining Containment Integrity l
B9.
Primary Containment Pressure (HC Category 1; j.
RG Category 11 Position:
Implemented.
See A6, B7, CS, C10, and D4.
B10. Primary Containment Isolation Valve Positior.
(excluding check valves) (HC Category 1; RG (a Secnco 6.2.4.2.)l Category 1)
Position:
Impl emen t ed/.
Redundant indication is not required on each redundant isolation valve.
1.8.1.97.3.3 Type C Variables l
E.
C1.
Radioactivity Concentration or Radiation Level in Circulating Primary Coolant (RG Category 1)
Position:
Not implemented.
See Issue 5, Section 1.8.1.97.4.5.
C2.
Analysis of Primary Coolant (gamma spectrum) (HC Category 3; RG Category 3)
Position:
Implemented.
C3.
BWR Core Thermocouples (RG Category 1)
Position:
Not implemented.
See B4, B5, and SLI-8121 (December, 1981) (Appendix A to Reference a.8-4).
t l.8-69 Amendment 7 J
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under maximum differential pr,essures, steam laden atmospheres, high temperature, high humidity, and radiation.
See Section 3.11 for further discussion of environmental qualification and Section 3.9.3 for further discussion of the operability of active components.
Generally, the containment isolation system is redundant and physically separated in its electrical and mechanical design, with diversity in parameters sensed for the initiation of containment isolation.
Power for the actuation of the two isolation valves in a line is supplied by two redundant, independent power sources without crossties.
In general, depending upon the system under consideration, the outboard and inboard containment isolation valves are powered and controlled by different electrical channels, with the supply source being Class lE ac f or both channels.
See Chapter 7 for further discussion of the control and instrumentation of the containment isolation system and Section 8.3 for a further discussion of onsite power systems.
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9MEb The containment isolation system is designed with provisions for administrative control, to ensure that the proper position of all nonpoweredfisolation valves is maintained.
All power-operated primary containment isolation valves /have position indicators in
/
cne main control room.
Discussion of instrumentation and yyjH controls for the isolation valves is included in Chapter 7.
9%mR NAllANE The design of the primary containment isolation system gives consideration to the possible adverse dynamic effects, such as water hammer, sudden isolation valve closure under normal operation, and to thermal expansion in those portions of pipe between the containment isolation valves.
The containment isolation system is designed so that failure of motive power is in the direction of greater safety.
Motor-operated isolation valves remain in their last position upon failure of electrical power to the motor operator.
Air-operated
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containment isolation valves are spring-loaded to close upon loss of air or electrical power to the pilot-operated solenoid valve.
Solenoid-operated isolation valves fail closed upon a loss of electrical power to the solenoid.
The 1.68 psig containment pressure setpoint that initiates containment isolation for nonessential penetrations is the minimum compatible with an acceptable plant availability for power production.
s 6.2-42 Amendment 12
Shutdown cooling suction valve BC-HV-F009 is powered fror Divisto I.
In the event that control of this valve is affected by tne fire, tnen an alternate shutdown path can be used.
The alternate shutdown cooling mode in lieu of RHR B shutdown cooling mode is as follows:
Core spray or LPCI D mode of RHR is used to fill up the reactor until the stear lines are flooded.
With one or more SRVs open, water flows out the relief valve and back
.c the suppression pool.
RHR B suppression pool cooling mode cools the suppression pool as before.
9A.S.3 Remote Snutdown Method Ine remote shutdown method can be founc' in the FSAR Appendix 9A
.k under response to NRC Generic Letter 81-12, Item 1.e.
In addition, depressurization by use of 3 SRVs and use of the B LPCI is available ~ rom the RSF.
9A.S.4 Scurious Sicnal Analvsis Resultr A complete review of spurious signals per section 9A.I.5.f was performed.
Tne circuits requiring separation were identified and
- fixes, i.e.,
cable rerouting, fire walls, etc., have been included in the plant design.
No cable tray fire wrapping is required and only one area of conduit fire wrap is required.
No wire cutting or fuse pulling operations are required.
The alternate shutdown mode was used in the reactor building and electrical access areas due to the logic and electrical channels associated with the RHR shutdown cooling valves.
These valves are BC-HV-F008, BC-HV-F009, BC-HV-F015A and BC-HV-F015B.
In addition, due to the large number of trips associated with the RCIC and HPCI, ma,nual depressurization and low pressure injection systems were relied on.
Because of the good cable and equipment separation at HCGS and the above shutdown methods, no manual actions are required to achieve hot or cold shutdown. 'These shutdown methods have been included in the plant operating procedures.
However, it may be desirable to manually establish RHR shutdown cooling or one of the high pressure injection systems to avoid the normally less desirable alternate shutdown path and/or fast depressurization to use LPCI or CS.
EEEd igh/ low prescure valve interf ace problems were identified as discussed in the response to Generic Letter 81-12, Item 2.b
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l 9A-32 Amendment 13 l
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2a.
I6'ntify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.
Response
The high-low pressure interfaces which use redundant electrically controlled devices to isolate the primary coolant boundaries are as follows:
^
1.
Reactor Vessel Vent Valves BB-HV-F001 BB-HV-F002 2.
RHR Suction Valves BC-HV-F00E BC-HV-F009 v
J 3.
...uValves BG-HV-F001 IN!ERTl BG-HV-F004 2b..,For each set of redundant valves identified in a.,
verify the redundant cabling (power and control) have adequate physical separation as required by Section III.G.2 of Appendix R.
Response
The required Appendix R,Section III.G.2, separation of the
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cabling for the above listed sets of valves cannot be provided.
The control wiring for each set of valves is separated using modified Regulatory Guide 1.75 separation in the remote shutdown.
panel, 10C399 and also in the main control room panelh'C650.i '1 2c.
For each case where adequate separation is not provided, show that fire-induced f ailures (hot shorts, open circuits, or shorts to ground) of the cables will not cause maloperation and result in a LOCA.
9A-94 Amendment 14 J
INSERT FOR PAGE 9A-94 BG-HV-F034 BG-HV-F035 4.
LPCI Injection / Bypass Valves BC-HV-F017A,B,C,D BC-HV-F146A,B,C,D 5.
RHR Return / Bypass Valves BC-HV-F015A,B BC-HV-F122A,B 6.
Core Spray Supply / Bypass Valves BE-HV-F005A,B BE-HV-F039A,B i
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Response
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Fire-induced f ailures of the cables f or the dette valves listed in[
item 2a.1,hanc e.;will not cause maloperation and result in acontroH jlLOCA since the valves are normally closed and(thcir T ;:r'M;c;; power is disconnected center y I nv r a c u na--out-t-ne-c r ea k e rs.; 9The Rh'CUivalvessare normally cpen pu RM and their cabling outside the control room, cable spread roome N
Di^*f EG-HU-and control equipment room have oeen verified to be separated by III.G.2.a separation.>A
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Therefore, a fire in the zones where the physical separation is less than Section III.G.2 requires will not cause maloperation and result in a LOCA.
See FSAR Section 7.4.1.4.
9A-95 Amendment 14 s
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INSERT FOR PAGE 9A-95 Therefore, for fires outside these areas, at least one valve will be available to isolate the RWCU system.
For a control / diesel area fire which might affect all 4 valves, spurious operation is prevented by removing power to the F034 and P035 valves during normal operation.
There are other valves asscciated with the filter /demineralizer portion of tne system which are not identified in Item 2a.3 above because they are located, powered and controlled in the reactor building and the separation of BG-HV-F001 and F004 is adequate to assure isolation from the main control room as discussed above (i.e.,
F001 and F004 are considered to be the high-low pressure interface for these valves).
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