ML20133P171
| ML20133P171 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/12/1985 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20133P165 | List: |
| References | |
| TAC-59439, TAC-59440, NUDOCS 8508140190 | |
| Download: ML20133P171 (23) | |
Text
, _ - - - _ - - _ _.. - - _ _ _ _. -.. -
'9 o 12.
(NPE-85/06)
SUBJECT:
. Technical-Specifications Tables 3.3.7.4-1, 3.4.3.2-1, 3.6.4-1, 3.8.4.1-1 and 3.8.4.2-1, pages 3/4 3-71, 3/4 4-11, 3/4 6-30, 3/4 6-31, 3/4 6-38, 3/4 8-32, and 3/4 8-47 DISCUSSION:
This technical specification change results in part from a design. change to add a more' accessible RHR to Head Spray containment isolation valve and is planned for completion during an outage currently scheduled to begin in October, 1985.
It is proposed to designate valve E12-F394 as the RHR to Head Spray Reactor Coolant System Pressure and Containment Isolation Valve and no longer classify valve E51-F066 as the inboard isolation valve. A footnote requiring any required revision of the proposed 35 second closure time
- be submitted within 90 days after com'pletion of the second ASME Section XI. closure time test is also proposed.
It is further proposed that valve E12-F344 be deleted from the Containment and Drywell Isolation Valve Table 3.6.4-1 and that the penetration protection circuit breaker and the thermal overload protection associated with the power circuitry for valve E12-F394 be added to tables 3.8.4.1-1 and 3.8.4.2-1, respectively.
'It is also proposed to make an editorial clarification change to ensure that the technical specifications reflect the correct nomenclature for the following valves:-
Page 3/4 3-71, Items 26 and 27; Reactor Core Isolation Cooling Test Return to Condensate Storage Tank Inboard, and Outboard,-Valves ("RCIC Test RTN to CST IB (OB) Valve"),
respectively.
Page 3/4 6-31, Valves E12-F028A-A and B-B; Residual Heat Removal Heat Exchanger "A" ("B") to Containment Spray Sparger Inlet ("RHR Heat Exchanger "A" ("B") to CTMT SPR Sparger INL"),
and for valves E12-F037A-A and B-B, "RHR Heat Exchanger "A"
("B") to CTMT Pool."
(It should be noted that attached technical specification page 3/4 6-31 reflects a previously requested change, NLS-85/06, which was transmitted as' Item 4 of MP&L serial AECM-85/0168 on July 3, 1985, and that attached technical specification page 3/4 4-11 reflects change NPE-85/15 which is requested by item 13 of this submittal.)
JUSTIFICATION:-
Local leak rate testing of the current RHR to Head Spray isolation valve E51-F066 requires removal of the drywell head and insulation because of the normally inaccessible location of the valve. To enhance operational effectiveness, a design change has been performed which added motor operated valve E12-F394 in a more accessible location upstream of valve E51-F066. The proposed technical specification change designates E12-F394 as the RHR to Head Spray isolation valve instead of E51-F066.
8508140190 850812 i
PDR.ADOCK 05000416 P
PDR J0P14 MISC 85042201 - 1
'J Since E12-F394 is upstream of the test connection which valve E12-F344 isolates, the test connection is no longer a leakage path from the drywell and it is also proposed to delete valve E12-F344 from table 3.6.4-1.
Valve E12-F394 does not have an analytical closure time defined in the FSAR accident analysis. The closure time for this valve must therefore be determined from ASME Section XI test data as described in Technical Specification Bases 3/4.6.4.
Since the valve cannot be made operable during power operation until af ter approval of this proposed change, this data is not yet available. The closure time is therefore qualified with a footnote that any required change to the closure time be submitted within 90 days after completion of the second ASME Section XI closure time test. The proposed closure time of 35 seconds was chosen as a reasonable interim closing time because valve G33-F253, which is of the same manufacture and was procured through very similar purchase specifications, is currently listed in the technical specifications and has been found to have a maximum closure time by technical specification bases criteria of 35 seconds.
It is expected that valve E12-F394 will exhibit closure characteristics similar to those of valve G33-F253.
The addition of the appropriate penetration protection circuit breaker and the thermal overload protection to tables 3.8.4.1-1 and 3.8.4.2-1 associated with valve E12-F394 will make the' technical specification consistent with the safety analysis and the as-built plant when the electrical connections are terminated af ter approval of this proposed change.
This design change as proposed adds a local handswitch on the remote shutdown panel for valve E12-F394. MP&L does not propose to add this handswitch to Technical Specification 3/4.3.7.4.
Neither the GGNS safety analysis nor the GGNS Fire Protection Plan requires operability of the E12-F394 valve from the remote shutdown panel for any analyzed condition.
It should be further noted that while the design change includes an interlock such as is presently included for valve E12-F023 to prevent inadvertant overpressurization of the RHR system through the head spray line, this interlock will not be included in the remote shutdown panel control circuit. Overpressurization of the RHR system through the head spray line from the remote shutdown panel would require the following sequence of events:
1.
A control room fire or similar circumstances which would render the main control room uninhabitable, thereby requiring operation from the remote shutdown panel, 2.
an operator error resulting in the opening of both motor operated valves E12-F023 and E12-F394 against a reactor pressure which is greater than the design pressure of the RHR system, and 3.
failure of the outboard check valve E12-F019.
The foregoing is considered to be an incredible sequence of events involving a minimum of three single failures, and incorporation of a pressure interlock for the remote shutdown system control circuit is therefore not required.
J0P14 MISC 85042201 - 2
...9 The changes correcting the valve nomenclatures are editorial changes proposed solely to enhance the clarity of the subject tables. They will ensure that the technical specifications reflect the specific functions of these valves and will make the technical specifications consistent with the specific nomenclature used in the as-built plant.
SIGNIFICANT HAZARDS CONSIDERATION:
The proposed changes are operational enhancements which will decrease operational down time during leak rate testing since it will no longer require removal of the drywell head and insulation to accomplish the testing. These changes will also reduce personnel exposure since radiation workers will not be required to remain in the radiation area for extended periods to accomplish the leak rate testing.
The addition of valve E12-F394 to the RHR Head Spray line and the installation of its attendant hardware were performed in accordance with~ applicable industry and regulatory codes and standards and the GGNS Quality Assurance Program, and as such are consistent with the licensing bases and the safety analyses. The additional changes requested by this item are purely administrative in nature in that they are corrections of valve nomenclature only. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident from any accident previously evaluated, nor do they involve a significant reduction in a margin of safety.
Therefore, the proposed changes involve no significant hazards considerations.
J0P14 MISC 85042201 - 3
- 3 E~~$.Sl~0f)
INSTRUMENTATION TABLE 3.3.7.4-1 (Continued)
REMOTE SHUTDOWN SYSTEM CONTROLS MINIMUM CHANNELS OPERABLE CONTROL Div 1 Div 2 b
b 12.
RHR Injection Valves 2
2 13.
RHR Test Line Valve 1
1 14.
1 15.
RHR HX Flow to Suppression Pool Valve 1
1 16.
RHR Discharge to Radwaste Valve 1
1 D
b 17.
2 18.
Diesel Generator HX Inlet Valve 1
1 b
b 19.
Safety / Relief Valves 6
6 20.
RHR to RCIC Head Spray Line Valve 1
NA 21.
RCIC Turbine Flow Controller 1
NA 22.
RCIC Suction Flow Suppression Pool Valve 1
NA 23.
RCIC Injection Shutoff Valve 1
NA 24.
NA 25.
RCIC Recirc. Main Flow Bypass Valve 1
NA RTN 26.
RCIC Test +GV to CST IB Vg/ye 1
NA OB 27.
NA A
28.
Steam to RCIC Turbine Valve 1
NA 29.
RCIC Turbine Trip & Throttle Valve 1
NA 30.
RCIC Turbine Cooling Water Valve 1
NA 31.
RCIC Turbine Local Control Select Switch 1
NA 32.
RCIC Gland Seal Compressor 1
NA 33.
Shutdown Cooling Isolation Valve Reset Switch 1 1
NOTE:
a.
1 per cooling tower fan b.
1 per valve GRAND GULF-UNIT 1 3/4 3-71 AmendecCE M8 + --
tz. (NPG-8s/06)
TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER SYSTEM E21-F005 LPCS E21-F006 E22-F004 HPCS E22-F005 E12-F008 RHR E12-F009 E12-F023 E12-F041 A, B, C E12-F042 A, B, C E12-F050 A, B E12-F053 A, B E12-F308
'E It~ F394 E51-F063 RCIC E51-F064 i
E51-F065 l
cr, en,,
E51-Fb76 E51-F013 TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES PRESSURE MONITORS - ALARM i
ALARM SETPOINT VALVE NUMBER SYSTEM (psig)
[
E21-F005 to E21-F006 LPCS 50-575 lp w
E12-F008 to E12-F006A RHR 5 183 ft)
E12-F008 to E12-F006B RHR 1 183 k
N E12-F041A to E12-F042A RHR 50- f75 g
v E12-F041B to E12-F042B RHR 1-60ll75 t-E12-F041C to E12-F042C RHR 1-6G-4 75 GRAND GULF-UNIT 1 3/4 4-11 Amendment No,
I2. (NPE -85/06.)
~', ' +
TABLE 3.6.4-1 CONTAINMENT AND ORYWELL ISOLATION VALVES MAXIMUM SYSTEM AND PENETRATION ISOLATION TIME VALVE NUMBER NUMBER VALVE GROUP (a)
(Seconds) 1.
Automatic Isolation Valves #
a.
Containment Main Steam Lines B21-F028A 5(0)*
1 5
Main Steam Lines B21-F022A 5(I)*
1 5
Main Steam Lines B21-F067A-A 5(0)*
1 9
Main Steam Lines B21-F0288 6(0)*
1 5
Main Steam Lines B21-F022B 6(I)*
1 5
Main Steam Lines B21-F067B-A 6(0)*
1 9
Main Steam Lines B21-F028C 7(0)*
1 5
Main Steam Lines B21-F022C 7(I)*
1 5
Main Steam Lines B21-F067C-A 7(0)*
1 9
Main Steam Lines B21-F0280 8(0)*
1 5
Main Steam Lines 821-F0220 8(I)*
1 5
Main Steam Lines B21-F0670-A 8(0)*
1 9
RHR Reactor E12-F008-A 14(0) 3 40 Shutdown Cooling Suction RHR Reactor E12-F009-B 14(I) 3 40 Shutdown Cooling Suction Steam Supply to E51-F063-8 17(I) 4 20 RHR and RCIC Turbine Steam Supply to E51-F064-A 17(0) 4 20 RHR and RCIC Turbine Steam Supply to E51-F076-B 17(I) 4 20 RHR and RCIC Turbine RHR to Head Spray E12-F023-A 18 3
94 RttR to dead Seray E12,-F39+-B 18 3
3 Main Steam Lihe B21-F019-A 19 1
20 Drains (a) See Specification 3.3.2, Table 3.3.2-1, for isolation signal (s) that operates each valve group.
(b) Deleted Hydrostaticallytestedbypressurizingsyst$m,12.65ps Hydrostatically tested with water to 1.10 P (c) to 1.10 P (d)
(e)
(f) Deleted (g) Normally closed or locked closed manual valves may be opened on an intermittent basis under administrative control.
- The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDIT0NS 2 or 3 provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching a reactor steam pressure of 600 psig and prior to entry into OPERATIONAL CONDITION 1.
- The "-A, -B, -C, -(A), -(B), -(C)" designators on the valve numbers indicate associated electrical divisions.
- Q'nSer9 GRAND GULF-UNIT 1 3/4 6-30 Amendment No. --
b
,1
/2... (NPE-85/06)
INSERT to page 3/4 6-30 Initial closure time. Pinal closure time to be determined during ASME Section XI testing. Any required change to this closure time shall be submitted to the Commission within 90 days of the second closure time test completion.
s i
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J0P14 MISC 85042201 - 4 L.
[
- 12. (NPE-95lc0 TABLE 3.6.4-1 (Continued)
CONTAINMENT AND ORYWELL ISOLATION VALVES MAXIMUM SYSTEM AND PENETRATION ISOLATION TIME VALVE NUMBER NUMBER VALVE GROUPsa)
(Seconds)
Containment (Continued)
Main Steam Line B21-F016-B 19(I) 1 20 Drains RHR Heat Exchanger E12-F028A-A 20(I) 5 90 "A" to kPC+ CTMT SPR Sporger ZNL RHR Heat Exchanger E12-F037A-A 20(I) 3 74 "A" to iPC+CTmT Peo/
l RHR Heat Exchanger ' E12-F028B-B 21(I) 5 90 "B" to -bSC4CTMT SPR Sperger %/fL l
RHR Heat Exchanger E12-F0378-B 21(I) 3 74 "B" to tPC4 CTMT Pool l
RHR "A" Test Line E12-F024A-A 23(0)(d) 5 90 to Supp. Pool RHR "A" Test Line E12-F011A-A 23(0)(d) 5 36 to Supp. Pool RHR "C" Test Line E12-F021-B 24(0)N 5
144
[
to Supp. Pool N
HPCS Test Line E22-F023-C 27(0)(d) 6B 75 h
RCIC Pump Suction E51-F031-A 28(0)(d) 4 56 RCIC Turbine E51-F077-A 29(0)(c) 9 26 Exhaust LPCS Test Line E21-F012-A 32(0) 5 144 Cont. Purge and M41-F011-(A) 34(0) 7 4
Vent Air Supply Cont. Purge and M41-F012-(B) 34(I) 7 4
4 Vent Air Supply Cont. Purge and M41-F034-(B) 35(I) 7 4
Vent Air Exh.
Cont. Purge and M41-F035-(A) 35(0) 7 4
Vent Air Exh.
Plant Service P44-F070-B 36(I) 6A 33 Water Return Plant Service P44-F069-A 36(0) 6A 33 Water Return Plant Service P44-F053-A 37(0) 6A 33
~
Water = Supply Chilled Water P71-F150-(A) 38(0) 6A 12 Supply i
GRAND GULF-UNIT 1 3/4 6-31
/?mendatertf N8 -
C 1
', Y.
/2,[N#6-$6/06)-
1 TABLE 3.6.4-1 (Continued)
CONTAINMENT AND DRYWELL ISOLATION VALVES SYSTEM AND PENETRATION
-VALVE NUMBER NUMBER b.
Drywell Cont. Cooling P42-F114-B 329(0)
Water Inlet Cont. Cooling P42-F116-A 330(I)
Water Outlet Cont. Cooling P42-F117-8 330(0)
Water Outlet 3.
Other Isolation Valves (9)#
a.
Containment Fuel Transfer F11-E015 4(I)
Tube Feedwater Inlet B21-F010A 9(1)
Feedwater Inlet B21-F032A 9(0)
Feedwater Inlet 821-F010B 10(I)
Feedwater Inlet B21-F032B 10(0)
RHR "A" Suction E12-F017A 11(0)(d) 12(0)((d)
RHR "B" Suction E12-F017B 13(0) d)
RHR "C" Suction E12-F017C RHR Shutdown E12-F308 14(I)
Cooling Suction
" : ' Z"" Z '
M G I! '"'
?!M(
' NR N'e't dE."I'A" E12-Fb44A 5b(5)
R a
to LPCI RHR Heat Ex. "A" E12-F025A 20(I) to LPCI RHR Heat Ex. "A" E12-F107A 20(I) to LPCI RHR Heat Ex. "B" E12-F025B 21(I) to LPCI RHR Heat Ex. "B" E12-F044B 21(I) to LPCI RHR Heat Ex. "B" E12-F107B 21(I) to LPCI RHR Heat Ex. "C" E12-F234 22(0) to LPCI RHR Pump "C" to E12-F041C-B 22(I)
LPCI RHR Pump "A" Test E12-F259 23(0)(')
Line to Suppr.
Pool RHR Pump "A" Test E12-F261 23(0)(')
Line to Suppr.
Pool GRAND GULF-UNIT 1 3/4 6-38 Amen 2Wat M8 -
(-
,'s l2e (NPE -8S/06)
TABLE 3.8.4.1-1 (Continued)
PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES c.
480 VAC Circuit Breakers (Continued)
Molded Case, Type NZM TRIP
RESPONSE
BREAKER SETPOINT TIME SYSTEM / COMPONENT NUMBER (Amperes)
(Seconds)
AFFECTED 52-1542-29 1200 0.100 STBY LIQ CONTROL SYS MIXING HEATER (Q1C41D003) 52-1611-10 12.5 O.100 MOV - DRYWELL COLL TK OUTLET ISOLATION (Q1G41F044-B) 52-1611-15 12.5 0.100 MOV - PSW CTMT STM TNL CLR ISOL (Q1P44F070-B) 50V-RNA Rx #p 52-gtt-/4 50
- o. goo
-g SPR ZNBD ZSQL 52-1611-25 12.5 0.100 MOV - DRYWELL
! ele /2F394 -Al)
CLG WTR ISOL (Q1P42F117-B) 52-1611-31 12.5 0.100 MOV - DRYWELL CLG WTR INL ISOL (Q1P42F114-B) 52-1611-32 32 0.100 MOV - CTMT CLG WTR ISOLATION (Q1P42F068-B) 52-1611-42 12.5 0.100 M0V PSW STEAM TUNNEL CLR ISOL (Q1P44F074-B) 52-1611-43 12.5 0.100 M0V PSW STEAM TUNNEL CLR ISOL (Q1P44F077-B) 52-1611-44 38 0.100 MOV - SERVICE AIR DRYWELL ISOLATION (Q1P52F195-B) 52-1621-03 7
0.100 MOV - DRWL HYDR INST LINE ISO (Q1E61F595B-B) 52-1621-04 7
0.100 MOV - DRWL HYDR INST LINE ISO (Q1E61F5978-B)
GRAND GULF-UNIT 1 3/4 8-32 Aasendment No.
w
[.'
o.
Ilo(NPE-$Slo TABLE 3.8.4.2-1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION
. BYPASS DEVICE (CON-TINU0US) (ACCIDENT SYSTEM (S)
VALVE NUMBER CONDITIONS) (NO)
AFFECTED Q1E51F010 Continuous RCIC System Q1E51F013 Continuous RCIC System Q1E51F019 Continuous RCIC System Q1E51F022 Continuous RCIC System Q1E51F031 Continuous RCIC System
-Q1E51F045 Continuous RCIC System Q1E51F046 Continuous RCIC System Q1E51F059 Continuous RCIC System Q1E51F068 Continuous RCIC System
.RCIC Trip and Throttle Continuous RCIC System Valve on Turbis.e Q1E51C002 Q1821F065A No Reactor Coolant System Q1821F065B No Reactor Coolant System Q1821F038A No Reactor Coolant System Q1B21F0988 No Reactor Coolant System
-Q1B21F0980 No Reactor Coolant System Q1821F098D No Reactor Coolant System Q1821F019 Continuous Reactor Coolant System
-Q1821F067A Continuous Reactor Coolant System
.Q1821F0678 Continuous Reactor Coolant System Q1821F067C Continuous Reactor Coolant System Q1821F0670 Continuous Reactor Coolant System Q1821F016 Continuous Reactor Coolant System
.Q1821F147A Continuous MSL Drain Post LOCA Leak-
~'
age Control Q1821F1478 Continuous MSL Drain Post LOCA Leak-age Control Q1833F019 Continuous Recirculation System
-Q1833F020 Continuous Recirculation System Q1833F125 Continuous Recirculation System Q1833F126 Continuous Recirculation System Q1833F127 Continuous Recirculation System Q1833F128 Continuous Recirculation System Q1023F591 Drywell Monitoring System Q1D23F592 Drywell Monitoring System Q1023F593 Drywell Monitoring System Q1D23F594 Drywell Monitoring System Q1E12F040 Continuous RHR System Q1E12F023 Continuous RHR System Q1E12F006A Continuous RHR System Q1E12F052A Continuous RHR System Q1E12F008 Continuous RHR System Qlal2FB49 C*MIwous RHA systent GRAND GULF-UNIT 1 3/4 8-47 Amend / net No -
- ', c
- p'"3.
j i
=13. (NPE-85/15)~
SUBJECT:
Technical Specification Tables 3.3.3-1, 3.3.3-2, 3.3.3-3, 4.3.3.1-1,. Technical Specification 4.4.3.2.2.b, Tables 3.4.3.2-2,
.and 3.4.3.2-3, pages 3/4 3-28, 30. -31, -33. -34, 3/4 4-10, -11, and -12 DISCUSSION:
This technical specification change results from a design change to add high/ low pressure interlocks to the injection valves on the low pressure ECCS systems and is planned for implementation-
'during an outage scheduled to begin in October, 1985.
It is proposed to add' appropriate requirements for the interlocks to.the subject specifications as follows:
1.
.Requfre three (3) minimum operable channels and action 31 for operational conditions 1, 2, and;3, and action 35 (proposed) for operational conditions 4* and 5*'(Table 3.3.3-1,-pages 3/4 3-28 and -30).
2.
Require a trip setpoint and allowable value of 516 psig,
~
decreasing and 452-534 psig, decreasing, respectively (Table 3.3.3-2, page 3/4 3-31).
3.
Delete ECCS System Response Times for LPCS and LPCI mode of RHR A, B and C (Table 3.3.3-3, items 1 and 2, page _3/4 3-33).
4.
Require ECCS actuation instrumentation surveillance intervals of once per:
twelve hours (S) for channel check, month (M) for channel functional test, and refueling (R(*)) for channel calibration in operational conditions 1, 2, 3, 4* and 5* (Table 4.3.3.1-1, page 3/4 3-34).
i:
5.
Delete the paragraph which implemented special requirements for the LPCS and LPCI check valves until completion of the proposed modification (Technical Specification 4.4.3.2.2.b, page 3/4 4-10).
6.
Revise the alarm and interlock setpoints from 50 to 575 psig for the LPCS' valve and from 50 to 475 psig for the LPCI valves (Tables 3.4.3.2-2 and -3, pages 3/4 4-11 and -12).
(It should be'noted that attached technical specification page 3/4 4-11 reflects change NPE-85/06 which is requested by item 12 of this 4
transmittal.)
c JUSTIFICATION: Facility Operating License NPF-29. Condition 2.C.(18), requires MP&L to implement isolation protection against overpressurization'of the low pressure emergency core cooling systems (RHR/LPCI and LPCS) prior to startup following the first refueling outage due to concerns about i
postulated intersystem LOCA. Completion of this design change will ensure compliance with the license condition.
J0P10PMI85052808 - 15 t
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's.,
The effect.of adding a pressure permissive interlock to the LPCS and LPCI systems was evaluated for the DBA recirculation suction line break with failure of the LPCI diesel generator. This break and failure combination was identified.in Section 6 of the GGNS FSAR-as
.the;1imiting event (LOCA). The DBA suction break is most sensitive to this design change because of its rapid depressurization rate.
Smaller. breaks depressurize more slowly and are less affected by-
~
addition.of the pressure permissive. The primary effect of the pressure permissive logic on this analyzed event is'to delay the time at which the-LPCS and LPCI systems begin injecting coolant into the reactor pressure vessel.
A new analysis of this postulated accident (using the existing model) was performed in which LPCI and LPCS injection was delayed by.the pressure permissive interlock. The new analysis, based on the conservative assumption that LPCI and LPCS flow does not begin until the injection valves are-fully open at approximately 58 seconds into the event, showed an increase in PCT of 51*F for a maximum PCT of
~
2149'F. Since the new PCT is still below the 10CFR50.46 limit of 2200*F, no change in the MAPLHCR limit is required.
The proposed addition of requirements to the technical specifications to support this design change will ensure appropriate surveillance off the valve interlocks. The changes result from analyses by the NSSS
~
-vendor, General Electric, and include adequate conservatism.to ensure the requirements will support continued safe operation.
Since the low pressure ECCS response times are now dependent on vessel depressurization time, which is different for each type of accident analysis, the technical specification definition of ECCS system response time is not applicable to these low pressure ECCS systems. The response time of 40 seconds is no longer applicable and should be deleted from the technical specifications. Diesel generator start' times (which are included in the present 40 second ECCS system response times) are tested by other technical.
j specification requirements. The maximum valve opening time of 30 seconds assumed in the analysis will be assured by the GCNS Inservice Testing Program.
The proposed minimum operable channels requirements of three per trip functics are applicable to the "one-out-of-two twice" logic incorpo-rated into the design change, and are adequate to ensure operability of the required low pressure injection functions considering the diversity of injection systems and logic channels available (four per trip function).
The proposed allowable value for the automatic interlocks includes an allowance for uncertainty and calibration inaccuracy and the proposed trip setpoint includes an allowance for instrument drift that will ensure the rated pressure of the low pressure piping will not be exceeded and that the assumptions used in the safety analysis will be met.
J10PMI85052808 - 16
r.
l
'S The proposed surveillance frequencies are consistent with those for-similar instrumentation already included in the technical specifi-cations and are adequate to ensure required availability of the instrumentation.
The special requirements for the LPCS and LPCI check valves incorporated in Specification 4.4.3.2.2.b are no longer required and should be deleted.
The revisions to the alarm setpoints in Table 3.4.3.2-2 and the interlocks to prevent manual opening of the valves in Table 3.4.3.2-3 are consistent with the additional proposed changes, and will serve to prevent inadvertent manual overpressurization of the low pressure piping from the normal control circuits. It should be noted that the manual opening interlock setpoint for LPCS is greater than the automatic pressure permissive setpoint. This is because the pressure permissive setpoint was established to meet the safety analysis assumptions, which assume a setpoint well below that required to protect the pressure integrity of the piping. The greater manual opening interlock setpoint is accept.ble to prevent exceeding the pressure rating of the LPCS piping during manual operation of the valves.
It should be further noted that consideration was given to including manual interlocks in the remote shutdown panel control circuits.
Overpressurization of these systems from the remote shutdown panel would require the following sequence of events:
1.
A control room fire or similar circumstances which would render the main control room uninhabitable, thereby requiring operation from the remote shutdown panel, 2.
an operator error resulting in the opening of the associated motor operated valves against a reactor pressure which is greater than the design pressure of the associated system, and 3.
failure of the associated system's outboard check valve.
The foregoing is considered to be an incredible sequence of events involving a minimum of three single failures, and incorporation of pressure interlocks for the remote shutdown system control circuits is therefore not required.
SIGNIFICANT IIAZARDS CONSIDERATION:
The design change will be performed in accordance with appropriate regulatory and industry codes and standards, the GGNS Quality Assurance Program and the applicable requirements of the GGNS FSAR.
The design change is therefore consistent with the licensing basis and the new safety analysis, which will be incorporated into the FSAR at the next annual update pending approval of these proposed changes. While the new analysis shows the design change will cause an increase of PCT to 2149'F during accident conditions, the 10 CFR 50.46 limit of 2200*F is not exceeded so the margin of safety is not reduced.
J0P10PMI85052808 - 17
-D o-.g 4
- The proposed technical specification changes that add requirements not presently included in the technical specifications are considered basically conservative changes. The change that deletes the ECCS response time requirements for these valves since they will no longer.
.tne applicable and revisions to existing applicable setpoints to make them consistent with the design change will make the affected technical specifications consistent with the plant as modified by the. proposed design changes.
It should also be noted that these changes are responsive to NRC concerns about the prevention of intersystem LOCA.
For the reasons cited, the proposed changes do not involve a significant' increase in the probability or consequences of an accident previously evaluated or create the possibility of a new or different kind of accident from any accident previously evaluated, not do they involve a significan* reduction in a margin of safety.
Therefore, the proposed changes involve no significant hazards considerations.
()
J0P10PMI85052808 - 18
e e9 Q
TABLE 3.3.3-1 E
o EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION O
MINIMUM OPERABLE APPLICABLE "n
TRIP FUNCTION (a)
CONDITIONS ACTION CHANNELS PER OPERATIONAL c$.
TRIP FUNCTION
--e A.
DIVISION I IRIP SYSTEM g
1.
RHR-A (LPCI MODE) & LPCS SYSTEM a.
. Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3, 4*, 5*
30 b.
Drywell Pressure - High 2(b) 1,2,3 30 c.
LPCI Pump A Start Time Delay Relay 1
1, 2, 3, 4 *, 5*
31 d.
Manual Initiation 1/ system (b) 1, 2, 3, 4*, 5*
32 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
a.
Reactor Vessel Water Level - Low Low Low, level 1 2
1,2,3 30 b.
Drywell Pressure - High 2
1,2,3 30 c.
ADS Timer 1
1,2,3 31 d.
Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 31 w
e.
LPCS Pump Discharge Pressure-High (Permissive) 2 1,2,3 31 x
f.
LPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 31 g.
Manual Initiation 2/ system 1, 2, 3 32 m
B.
DIVISION 2 TRIP SYSTEM I.
Reactor Vessel Water Level - Low, Low Low, Level 1 2
1, 2, 3, 4 *, 5*
30 b.
Drywell Pressure - High 2
1,2,3 30 c.
LPCI Pump B Start Time Delay Relay 1
1, 2, 3, 4 *, 5*
31 d.
Manual Initiation 1/ system (b) 1, 2, 3, 4*, 5*
32 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
l a.
Reactor Vessel Water Level - Low Low Low, Level 1 2
1,2,3 30 b.
Drywell Pressure - High 2
1,2,3 30 N
g c.
ADS Timer 1
1,2,3 31 W
d.
Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 31 g
,g LPCI Pump B and C Discharge Pressure - High (Permissive) 2/ pump 1, 2, 3 31 g
e.
- =
f.
Manual Initiation 2/ system 1, 2, 3 32 A.
g g}
m I
e4 U$
g
~
- e. encs., vessa erasgae -t- (x+cs A :ss) 3
,, 2, 5 3:
4*, 5
- 35 6
13.
(NPE-85/15)
INSTRUMENTATION TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
a.
With one channel inoperable, place the inoperable channel in the tripped condition within one hour
- or declare the associated system (s) inoperable.
b.
With more than one channel inoperable, declare the associated system (s) inoperable.
ACTION 31 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.
ACTION 32 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated ADS trip system or ECCS inoperable.
ACTION 33 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within one hour
- or declare the HPCS system inoperable.
ACTION 34 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour
- or declare the HPCS system inoperable.
--->~
"The provisions of Specification 3.0.4 are not applicable.
5CTION35-With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within one hour
- or declare the associated system (s) inoperable.
GRAND GULF-UNIT 1 3/4 3-30 Amendment No. ___
y."
TABLE 3.3.3-2 cs g
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS E
.ALLOWA8LE TRIP FUNCTION TRIP SETPOINT VALUE 7,
A.
DIVISION 1 TRIP SYSTEM g
1.
RHR-A (LPCI MDDE) Al.D LPCS SYSTEM Q
a.
Reactor Vessel Water Level - Low Low Low, Level 1
,-150.3 inches *
> -152.5 inches
-)
b.
Drywell Pressure - High 5 1.39 psig
$ 1,44 psig g
c.
LPCI Pump A Start Time Delay Relay 5 5 seconds
< 5.25 seconds t d.
Manual Initiation NA HA 2." AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" a.
Reactor. Vessel Water Level - Low Low Low, Level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 31.39psig 31.44psig c.
ADS Timer
< 105 seconds
< 117 seconds d.
Reactor. Vessel Water Level-Low, level 3 5 11.4 inches
- F 10.8 inches e.
LPCS Pump Discharge Pressure-High 145 psig, increasing 125-165 psig, increasing f.
LPCI Pump A Discharge Pressure-High 125 psig, increasing-115-135 psig, increasing R*
g.
Manual Initiation MA NA a
B.
DIVISION 2 TRIP SYSTEM wg 1.
Reactor Vessel Water Level - Low Low Low, Level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 31.39psig 31.44psig c.
LPCI Pump B Start Time Delay Relay 5 5 seconds
< 5.25 seconds u d.
Manual Initiation NA HA 2." AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" l
a.
Reactor Vessel Water Level - Low Low Low, Level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 31.39psig 31.44psig g
c.
ADS Timer
< 105 seconds
< 117 seconds f
d.
Reactor Vessel Water Level-Low, Level 3 5 11.4 inches
- i 10.8 inches
'Q LPCI Pump B and.C Discharge Pressure-High 125 psig, increasing 115-135 psig, increasing m,
3 e.
(
f.
Manual Initiation NA NA 3
C.
DIVISION 3 TRIP SYSTEM
'g%
1.
HPC5 SYSTEM
-f a.
Reactor Vessel Water Level - Low Low, Level 2
>-41.6 inches *
>-43.8 inches N
- b. Drywell Pressure - High 31.39psig
.31.44psig k
b c.
Reactor Vessel Water Level - High, Level 8
< 53.5 inches *
< 55.7 inches b
d.
Condensate Storage Tank Level - Low i 0 inches F -3 inches 5
e.
Suppression Pool Water Level - High 7 5.9 inches 7 7.0 inches f.
Manual Initiation
~
HA NA.
- e. Reache rassel Pressare Low 5% M*,h decreasl9 452-53y decreasjy (xnje& *en tem'ssive) p,
l3,(NfE~85l5
.,. r z, a
TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES (SECONDS) 1.
LOWPRESSURECORESPRNYSYSTEM 100 NA 2.
LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM PUMPS A, B AND C -4G-NA 3.
AUTOMATIC DEPRESSURIZATION SYSTEM NA 4.
HIGH PRESSURE CORE SPRAY SYSTEM
< 27 5.
LOSS OF POWER NA GRAND GULF-UNIT 1 3/4 3-33 8 *b*
s TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS "z
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH G:
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED A.
DIVISION I TRIP SYSTEM i
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM a.
Low Low Low, Level 1 5
M R
1, 2, 3, 4*, 5*
b.
Drywell Pressure - High S
M R
1,2,3 c.
LPCI Pump A Start Time Delay Relay NA M(b)
Q 1, 2, 3, 4*, 5*
d.
Manual Initiation NA R
Q 1, 2, 3, 4*, 5*
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
R a.
R(g,))
3 1, 2, 3 Low Low Low, Level 1 S
M 1
Y b.
Drywell Pressure-High 5
M R
1,2,3 y
c.
ADS Timer NA M
Q 1,2,3 d.
R(a) 1, 2, 3 Low, Level 3 S
M e.
LPCS Pump Discharge R(a) 1,2,3 Prassure-High S
M f.
LPCI Pump A Discharge R(a) 1,2,3 Pressure-High S
M(b) g.
Manual Initiation NA R
NA 1,2,3 W
B.
DIVISION 2 TRIP SYSTEM Q
l 1.
D b
a.
}
Low Low Low, Level 1 S
M R
1, 2, 3, 4*, 5*
f x
b.
Drywell Pressure - High 5
M R
1,2,3 gq c.
LPCI Pump B Start Time Delay Relay NA M(b)
Q 1, 2, 3, 4*, 5*
h d.
Manual Initiation NA R
Q 1, 2, 3, 4*, 5*
g I
A. Reactor Vassal Pressar< sow S W
,g g
g,1, y, y t gg
_ (rnyttk Perm.ssuve)
.. < C *...
/3,(pre-es/s)
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1_ The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
a.
Monitoring the drywell atmospheric particulate and gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
. Monitoring the drywell floor and equipment drain sump level and flow rate at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, c.
Monitoring the drywell air coolers condensate flow rate at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and d.
Monitoring the reactor vessel head flange leak detection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
a.
At least once per 18 months, and b.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.
Q edditien, until the L"CS ysic;;; :nd the RMR cy;ter injection ":'"e rea coolant pressure-low permissive is modified during or befo the first ling outage, the LPCS system check valve 1E2 -
and the RHR system c alves 1E12-F041 A, B, and C sh so be demonstrated OPERABLE by f ing leakage to ithin its limit:
1.
Whenever the unit has been
- TDOWN or REFUELING, after the last valve disturban ior to reactor nt system temperature exceeding 200*
2.
n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following valve disturbance except when in (HiiTnnww nr arrligtIq The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
4.4.3.2.3 The high/ low pressure interface valves leakage pressure monitors shall be demonstrated OPERABLE with alarm and interlock setpoints per Table 3.4.3.2-2 and Table 3.4.3.2-3 by performance of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 31 days, and b.
CHANNEL CALIBRATION at least once per 18 months.
An@b"
'~
GRAND GULF-UNIT 1 3/4 4-10
. e :*..
t 3. Ops-es/is) i TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER SYSTEM E21-F005 LPCS E21-F006 E22-F004 HPCS E22-F005 E12-F008 RHR E12-F009 E12-F023 E12-F041 A, B, C E12-F042 A, B, C E12-F050 A, B E12-F053 A, B E12-F308 E/2-F3 W E51-F063 RCIC i
E51-F064 00 E51-F065
-Efrt-F066 l
E51-F076 v
E51-F013 TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES PRESSURE MONITORS - ALARM ALARM SETPOINT VALVE NUMBER SYSTEM (psig)
E21-F005 to E21-F006 LPCS
-+ 675 E12-F008 to E12-F006A RHR 1 183 E12-F008 to E12-F006B RHR 1 183 E12-F041A to E12-F042A RHR 50 //-75 E12-F041B to E12-F042B RHR 50-lh75 E12-F041C to E12-F042C RHR 50 l/ 75 GRAND GULF-UNIT 1 3/4 4-11 Apeulmed MC
- 13. (NAE-2fl)S)
TABLE 3.4.3.2-3 REACTOR COOLANT SYSTEM INTERFACE VALVES PRESSURE INTERLOCKS INTERLOCK SETPOINT VALVE NUMBER SYSTEM (psia)
E12-F052 to E51-F064 RCIC 5 465 E12-F041A to E12-F042A RHR 50 l/ 75" E12-F041B to E12-F042B RHR i 50 F7f E12-F041C to E12-F042C RHR
-$-50 //75 E21-F005 to E21-F006 LPCS 50-575 GRAND GULF-UNIT 1 3/4 4-12 AntenJM8t
- ' "