ML19257D553
| ML19257D553 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 12/29/1979 |
| From: | Kelley J NORTHEAST UTILITIES |
| To: | |
| Shared Package | |
| ML19257D550 | List: |
| References | |
| NUDOCS 8002040596 | |
| Download: ML19257D553 (15) | |
Text
/- I - / /
APP OVAL h N ~ % DATE:
Plant Superintende..c STATIO!! PROCEDURE COVER SilEET A. IDEtiTIFICATIO:1 flumber OP 2509 Rev. 6 Title STEAM LIfiE RUPTURE Prepared By J. Kelley B. REVIEW _
I have reviewed the above procedure ard have found it to be satisfactory.
TITLE SIG!1ATURE DATE DEPARTf1EtlT l'EAD jQ ggeg/pp U $V C. UlREVIEtlED SAFETY OUESTI0ri EVALUATIO!! 00CU:4EllTATIO1 REQUIRED:
(Significant change in procedure method or scope as described in FSAR) YES [ ]
fl0 ((
(If yes, document in PORC/SORC meeting minutes)
EllVIR0i1ME 1TAL IfiPACT (Adverse environmental impact)
(If yes, document in PORC/SORC meeting minutes)
YES [ ] fl0 [8 D. PORC/SORC APPROVAL, PORC/SORC Meeting flumber 19 -l'M E. APPROVAL AflD If*Tu.EMEllTATION -
The attached procedure is her,eby approved, and effective on the dates below: -
/ Eff9 9 / ?O Plant Superintendent / Unit Superintendent Approved Date Effective Date SF- 301 I Rev. 2
'1867 224 8002040 5+%6
OP 09 Page 1 Rev. 6 Date: 1/1/80 STEAM LINE RUPTURE PAGE NO. EFF. REV. DATE 1 6 1/1/80 2 6 1/1/80 3 6 1/1/80 4 6 1/1/80 5 6 1/1/80 6 6 1/1/80 7 6 1/1/80 8 6 1/1/80 9 6 1/1/80 10 6 1/1/80 11 6 1/1/80 12 6 1/1/80 13 6 1/1/80 14 6 1/1/80 1867 225
OP 2. Page 2 Rev. 6 Date: 1/1/80
- 1. OBJECTIVE To provide a procedure which will limit the expected rapid cooldown of the reactor coolant system (RCS) when steam flow is excessive due to a steam line rupture or rupture of a feedwater line downstream of the air assisted check valve.
- 2. DISCUSSION Excessive steam flow from the steam generators (S/G) results in a cooldown of the RCS. This adds positive reactivity to the reactor (due to the negative moderator temperature coefficient) which results in a decrease in shutdown margin. The main steam isolation valves (MSIV) minimize the consequences of excess steam flow for breaks either upstream or downstream of the MSIV's.
If a steam line rupture occurs upstream of the MSIV's, the reverse flow check valve feature of the MSIV's will prevent a reverse flow of steam from the unaffected S/G. When pressure in the affected S/G decreases to 500 psia, the following will occur.
- a. Both MSIV's will trip closed.
- b. Both feedwater control valves close.
- c. Both air assisted feedwater check valves trip closed.
- d. Both feedwater pumps trip.
- e. n reactor trip signal is generated by the reactor pro-tection system (RPS).
The S/G that has suffered the rupture will, boil dry, stopping all heat transfer in that loop and therefore tenninating the cooldown of the RCS. ,
The unaffected S/G and auxiliary feedwater system are utilized to bring the unit to a cold shutdown condition using OP 2205 (Plant Shutdown) and OP 2207 (Plant Cooldown) as guidelines.
1867 226
OP 2L Page 3 Rev. 6 Date: 1/1/80 l If a feedwater line rupture occurs downstream of the air assisted check valve, steam will flow out of the break and the end result will be very similar to steam line break upstream of the MSIV's.
The resultant cooldown of the RCS will be less severe for the folloning reasons:
- a. Feedwater line are smaller than main steam lines (18 inches versus 34 inches).
If a steam line rupture occurs downstream of the MSIV's, both MSIV's will close when the pressure in either S/G falls to 500 psia. Excessive cooldown of the RCS is teminated and both S/G are unaffected and available for plant cooldown.
During a steam / feed line break inside containment, elevated containment temperatures will cause erroneous pressurizer and steam generator level indication. Procedure steps specifying control points based on level take necessary corrections into consideration. For containment temperatures up to 300 F the pressurizer water level will be between the taps if indicated level is between 19% and 87%. For S/G's the level must be between 15% and 90%.
Due to the similarity of symptoms for a loss-of-coolant inc. dent (LOCI) and a steam line rupture, it is difficult to differentiate between the two events. There are certain parameters which can be used to accurately distinguish which of the following has occurred:
- a. Loss-of-coolant incident (LOC:)
- b. Steam line rupture downstream of the MSIV's
- c. Steam line rupture upstream of the MSIV's/feedwater line rupture downstrna of air assisted check valve and outside the containment.
1867 227
OP 2 Page 4 Rev. 6 Date: 1/1/80
- d. Steam line rupture /feedwater line rupture inside the containment.
The block diagram on Figure 7.1 provides the operator with a means of establishing which emergency is occurring. In order to accomplish '
this identification, the first four steps of this procedure and Emergency Procedure 2506, loss-of-Coolant Incident (LOCI), are identical.
NOTE: Small steam line breaks that do not result in closure of the MSIV's or reactor trip are not considered an emergency.
- 3. SYl1PT0f15 NOTE: All symptoms may not initially be present. Any single or multiple symptom which provide an indication of a steam line rupture will require implementation of this procedure.
3.1 Major Symptoms 3.1.1 Major symptoms of steam line rupture upstream of MSIV's/feedwater line rupture downstrean of air assisted check valve.
3.1.1.1 Pressure in one S/G will decrease rapidly while pressure in the other S/G will be erratic until stabilizing at the pressure corresponding to the new Tavg (C05).
3.1.1.2 MSIV's will trip closed when the pressure decreases to 500 psia in the affected S/G (C05). --
3.1.1.3 Pressure in the affected S/G will continue to decrease, while the pressure i'n the unaffected S/G will decrease to the satura-tion pressure corresponding to the existing Tavg resulting from the cooldown (C05).
3.1.1.4 If rupture is inside containment, containment temperature and pressure will rapidly increase (C01).
3.1. 2 Major symptoms of steam line rupture downstream of MSIV's.
1867 228
OP 2. s Page 5 Rev. 6 Date: 1/1/80 3.1. 2.1 Pressure will decrease rapidly in both S/G's(C05).
3.1. 2. 2 MSIV will trip closed when the pressure decreases to 500 psia in either S/G (C05).
3.1. 2. 3 Pressure in both S/G's will stabilize and slowly return to the saturation pressure corresponding to the existing Tavg, approxi-mately 532 F, which is being controlled by the atmospheric dump valves (C05).
3.2 Other 3.2.1 Reactor trip from low steam generutor pressure, high roactor power and/or thermal margin / low pressure (C04).
3.2.2 Rapidly decreasing pressurizer level and pressure (C05).
3.2.3 Mismatch between steam flow /feedwater flow (C05).
3.2.4 SIAS initiation (C01).
3.2.5 Containment isolation (C05).
3.2.6 High noise level.
- 4. AUT0f1ATIC ACTIONS
_. 4.1 See verifications in Sections 5 and 6.
- 5. IMMEDIATE ACTION 5.1 Carry out the Immediate Actions of Emergency Procedure 2502 (Emergency Shutdown).
5.2 Five seconds after identifying CEA's fully inserted and verifying a low pressurizer pressure safety injection acutation trip all four (4) reactor coolant pumps (C03).
5.3 If pressurizer pressure is less than 1600 psia and SIAS has not initiated, nanually initiate SIAS, CIAS and EBFAS (C01).
1867 229
OP 2 u Page 6 Rev. 6 Date: 1/1/80 5.4 If steam generator pressure is less than 500 psig, verify auto l closure, or manually close, the main steam isolation valves (MSIV's) (C05).
5.5 Check Panel C01X to verify that all equipment associated with l the following signals is in the accident mode.
Safety injection actuation signal (SIAS).
Containment isolation actuation signal (CIAS). l Enclosure building filtratioa actuation signal (EBFAS).
5.6 As pressurizer pressure decreases, verify the following:
5.6.1 High pressure safety injection (HPSI) flow starts at approximately 1100 psig (C01).
5.6.2 Low pressure safety injection (LPSI) flow starts at approximately 200 psig (C01).
5.6.3 Safety injection (SI) tank level and pressure start decreasing at approximately 200 psig (C01).
5.7 Check containment pressure (C01):
5.7.1 Verify containment isolation at a containnent pressure of 5 psig or pressurizer pressure at 1600 psia.
(C0lX).
5.7.2 Verify containment spray actuation at 27 psig (C0lX).
5.8 Initiate emergency plan (Emergency Procedure 2501). !
5.9 Monitor quench tank level, pressure and temperature, PORV acoustic position indication and PORV discharge pipe temperature.
If a stuck PORV is indicated (RCS pressure < 2300 psia) close the blocking valve of the stuck / leaking PORV (C03).
- 6. SUBSEQUENT ACTIONS 6.1 Refer to Figure 7.1 to identify the emergency. If the emergency cannot be clearly identified assume a LOCA and refer to E0P 2506. If a LOCA or steam generator tube rupture is indicated refer to E0P 2506/2515.
6.2 Secure all feedwater to the affected steam generator (C05).
1867 230
e .
OP 2 Page 7 Rev. 6 Date: 1/1/80 NOTE: k:henever hot leg temperature (Th) is specified, the incore thermocouple temperature should also be monitored. This will be particularly useful if Th is off-scale. To demand incore thennoccuple reports fran the computer, enter "X SP 1 SP 0 E0T."
6.3 If not already closed, shut the main steam isolation valves (C05).
6.4 While pressurizer pressure and level control have not been reestablished.
6.4.1 Verify that the steam flow out the break, or the atmospheric dump valves if the break has been isolated, are maintaining T at saturated conditions for the H
pressurizer pressure (C05/C03).
6.4.2 Continue feeding the unaffected steam generator and return level to between 70% and 80% with the following constraints (C05).
6.4.2.1 Do not cause a RCS cooldown in excess of 100 F/ hour.
6.4.2.2 Do not feed the steam generator if RCS temperature is less than 400 F and decreasing.
6.4.2.3 Do not violate the water hammer limits of 600 GPM per steam generator when level is less than 45% or 168 GPM per steam generator when level is less than 45% and all feed flow has been lost for greater th 115 -
minutes.
6.4.3 Verify natural circulation has initiated and an RCS AT of greater than 10 F is indicated (C03).
6.4.4 Ensure continued required injection by the charging, HPSI and LPSI pumps (C01/C03).
6.4.5 Monitor voiding in the core by observing RCS AT and excore nuclear indication. If natural circulation appears to be degraded or lost (RCS AT >50 F) complete the following:
1867 231
OP 2 Page 8 Rev. 6 Date: 1/1/80 6.4.5.1 Af ter reverifying the absence of a LOCA, attempt to restart any RCP deemed operable per OP 2301C (C03).
6.4.5.2 If RCP's are not available ensure T H is maintained saturated by boiling / reflux in the steam generators (C03).
6.5 Verify actuation of containment spray if containnent pressure reaches 27 psig (C0lX).
I 6.6 Initiate naximum S/G blowdown on affected S/G (C05).
6.7 Uhen the pressurizer is reflooded using either of the following l methods, regain pressurizer pressure control .
6.7.1 When level is restored to between 40% and 50%, l energize pressurizer heaters and reestablish nomal pressure control (C03) or 6.7.2 Using HPSI pumps and charging pumps, go solid and control pressure by securing pumps (see Step 6.12) and throttling the safety injection valves (C01/C03).
NOTE: Use the subcooling meter only when Th > 515 F.
For lower temperatures refer to Figure 7.1 to detennine subcooling margin.
6.8 When pressurizer pressure control is returned establish at l least a minimun 50 F subcooling margin by coolirg the RCS or raising pressurizer pressure (C05/C03).
6.9 Verify the elimination of voids in the RCS by observing proper l RCP operation ,(C03). _
6.9.1 Pump current: greater than 400 amps 6.9.2 Steam Generator AP: >10 psid (2 p;mps)
>20 psid (3 pimps)
>30 psid (4 p;mps) 6.9.3 RCS aT: <5F 6.9.4 RCP vibration: Lack of alam 1867 232
OP J9 Page 9 Rev. 6 Date: 1/1/80 6.10 Determine any adverse environnental effects to the following l control equipment, depending on steam / feed line break location.
NOTE: Environmental conditions, depending on the steam / feed line break location, nay result in erroneous indications on instrumentation or equipment which are not environmentally qualified. For this reason, diverse means of verifying the below listed conditions in conjunction with .nonitoring environmentally qualified instrumentation, where provided, is required.
6.10.1 Verify the status of pressurizer power operated l relief valves (steam / feed line break inside containment) by acoustic valve position, block valve position l discharge tenperature indication, quench tank pressure /
temperature / level, pressurizer pressure and pressurizer level (control channels are environmentally qualified).
If one or both PORV's indicate stuck open (RCS pressure less than 2300 psia), proceed as follows:
6.10.1.1 Close both blocking valves (2-RC-403/405).
6.10.1.2 If either blocking valve does not indicate closed either by valve position or continued indication of PORV flow, secure power to both PORV solenoids bt opening breaker #3 on 125 VDC Panel 201 A-2 and breaker #3 on 125 VDC Panel 2018-2.
6.10.2 Verify the status of the atmospheric dump valves (steam / feed line break outside containment) to ensure the valve on the intact steam generator is closed by valve position, steam generator pressure and level, and visual indication of steam discharge.
If the atmospheric dump valve on the intact steam generator indicates open, proceed as follows:
1867 233
\
i OP 25b., Page 10 Rev. 6 Date: 1/1/80 I
i 6.10.2.1 Secure power to . .e affected valve pressure control loop by placing the atmospheric ;
dump quick-open handswitch HS-4224-C05) i in " disable", and opening the knife switch located in C05R, "TDQ" for the A valve and "TBM" for the B valve.
6.10.2.2 If the intact steam generator atmospheric dump valve still indicates open, secure instrument air to the Enclosure Building by shutting 2-IA-29.
NOTE: The following step is accomplished to prevent contain-ment equipment damage due to unnecessary containment spray.
6.11 Reset the containment spray actuation modules and secure the containment spray pumps if all the following conditions exist.
(C01/ local) .
The affected steam generator pressure is less than j 6.11.1 50 psig (C05).
6.11.2 Containment pressure is less than 10 psig (C01).
6.11.3 The pump is not operating on containment sump recircu-lation (C01).
6.12 Af ter carefully evaluating the indicated parameters, secure i the specified equipment only if other abnomal conditions require.
NOTE: When securing any equipnent, if possible, first ,
reset the actuation module on the safeguard panel .
If after securing the equipment conditions worsen to the point where the prerequisites are no longer met, .
restart the secured equipment.
6.12.1 Secure HPSI pumps only when all the following condi- l tions exist (C01). ,
6.12.1.1 The HPSI pumps suction is still aligned to j ,
the RWST (no SRAS)(C01). l' I
l i
1867 234
OP 9 Page 11 Rev. 6 Date: 1/1/80 6.12.1.2 All RCS hot leg and cold leg temperatures !
are at least 50 F subcooled (C03).
6.12.1.3 Pressurizer level has been restored and l responds to HPSI/ charging pump operation.
i (C03).
6.12.1.4 The reactor is shutdown as indicated by f all rods inserted and cold shutdown boron concentration (by sample).
(C04).
6.12.1.5 Core cooling is being provided by the !
steam generators or shutdovm cooling.
For steam generator opera' fon forced flow or natural circulation (Tn - T c< 50 F) is l required and the steam gene:ators must have steam and feed flow and level between 70% and 80% (C01/5).
6.12.2 Secure HPSI pumps if minimum pressure temperature l limits as specified in technical specffications are approached (T < 275 F)(C0 3/1). (This step is to be c
completed independent of conditions sp!cified in 6.12.1).
6.12.3 Secure the charging pumps if Th is greter than 50 F l below saturation temperature (Figure 7/.2), and pressurizer level is restored and respnds to charging i pump operation and , pressurizer _ pressum is under ,
control (C02).
6.12.4 Secure other safeguard equipment only after detemining other equipment is providing the desied function or the function is no longer required.
Final Condition - pressurizer pressure and level control restored, T at least 50 F suttooled, prepared H
to start RCS cooldown.
1867 235
/ r OP J9 Page 12 Rev . 6 Date: 1/1/80 l
- 7. FIGURES 7.1 Break Identification Chart 7.2 Saturation Curve O
s ,
OP 2509 l' age 13 ,
Rev. 6 Date: 1/1/80
. . ..-..2 .. s-BREAK IDE!!TIFICATICil CllAP,T . ,
Per Level Changing ,. . .
~ - .
- and . .
Pzr. Press. Rapidly Decreasing ,
N bl -
~ ~
Pressure .
L .
Abnormally Low In One '
flormal Or ,
Or Both Steam Generators Rising 4 1 LOCA Steam Line Break '
& 4 Containment Pressure Containment Pressure -
IL \'
/ b 11ormal UP liormal UP .
& 4 4 d, 145/FW Break
. LOCA S/G "ube. .
145/FW Break -
Out of CTriT
' ~
Rupture.-
In CT14T ,
Y - V Refer to s/
Pefer to Refe\/
r to Refer to ' '
EP 2506 , EP 2515 EP 2509 . EP 2509
^
1867 237
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