ML20212P797
| ML20212P797 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/02/1987 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20212P471 | List: |
| References | |
| C.12, TAC-64896, NUDOCS 8703160269 | |
| Download: ML20212P797 (11) | |
Text
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.c EFFECTIVE DATE 03-02-87 Rev. 5
/,,'- WP1852P i 0-0054P j C.12 LOSS OF DECAY HEAT REMGVAL SYSTEM 1.0 SYMPTOMS '
1.1 "RC DH CLR A(B) Flow Lo" Alarm i
1.2 "DH PMP A(B) Suct Hdr Temp Hi" Alarm 1:
I 1.3 "DH CLR A(B) Flow Hi-Lo" Alarm 1.4- "4160 V SWGR BUS 4A(B) FDR Trip", Alarm 1.5 Abnormal flow indication for decay heat cooler outlet flow 1.6 Increasing reactor coolant temperature 1.7- Decreasing pressurizer, fuel transfer canal or reactor vessel level.
{ 2.0 RESPONSE 2.1 Low flow Condition 2.1 .1 Check on line decay heat pump not running. Attempt restart if no annunciator on H2ES "4160 V SWGR bus 4A(B) FDR trio".
2.1 . 2 Check suction valves HV-20001 and HV-20002 open, check pump suct!on valves HV-20005 and/or HV-20006 open.
2.1 3 Check DH cooler A(B) discharge and bypass valves open SFV-26039 and/or SFV-26040, HV-26037 and/or HV-26038 open.
2.1 4 Check discharge valves SFV-26005 and/or SFV-26006 open.
2.1 . 5 Check RB sump valves HV-26105 and/or HV-26106 closed.
2.1 .6 Line up and place alternate decay heat system loop in service per OP A.8.
'i 8703160269 870310 PDR ADOCK 05000312 P PDR Rev. 5 C.12-1
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RESPONSE (Continued)
Ts' 2.1 .7 E both decay heat loops are inoperable, with RCS filled, vented and steam bubble in PZR TH Q:
2.1 .7.1 Notify the Unit Operation Superintendent and Nuclear Operation Manager AS SOON AS POSSIBLE.
2.1 .7.2 '
Monitor incore thermocouples by using either:
- 1. IDADS Computer
- 2. SPDS ,
4
- 3. Bailey Computer 1
2.1 .7.3 Maintain decay heat removal using the OTSEs as follows:
5* 2.1 .7.3.1
- E condenser vacuum is established. THEN open T8Vs as recessary to control RCS temperature.
2.1 .7.3.2 Notify Chemistry to sample steam as soon as possible to '
evaluate any possible off-site releases due to potential primary to secondary leakage.
54 2.1 .7.3.3 E the condenser is unavailable to accept steam, THEN Open
- ADVs (remotely or locally) as necessary to control RCS temperature.
2.1' .7.4 Maximize RCS letdown flow, consistent with availab.le f'
- makeup. This will increase RCS heat removal by the Letdown System.
5+ 2.1 .7.5 Verify conditions and support systems are established as
- necessary for running RCPs per A.2 " Reactor Coolant Pump System" in conjunction with subsequent steps.
2.1 .7.6
! Make every attempt to regain Decay Heat Pump and Decay Heat Flow.
2.1 .7.7 Monitor and maintain 50* to 80' RCS subcooling margin.
.N_qTE,:
The next steps performed verify establishment of conditions necessary for the OTSGs to remove heat from the RCS.
5 ++ 2.1 .7.8 E incore thermocouples reach 200* THEN:
l 2.1 .7.8.1
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Verify Open TBVs, E not available THEN:
, 5++ 2.1 .7.8.2 Open ADVs from Control Room, E not available THEN:
i Rev. 5 C.12-2 '
i
F-s 4' j RESPONSE (Continued) r"s s' 2.1 .7.8.3 Manually open ADVs, IE not available THEN:
63 2.1 .7.8.4 Establish condenser vacuum. IF NOT possible, THEN, 54 2.1 .7.8.5 Provide condenser cooling with MCW or PCW system AND Open HV-35033 and HV-35034, Condenser Vacuum Breakers, to provide alternate steam flow path. ,
- 2.1 .7.8.6 Open bypass around T8Vs.
54 2.1 .7.9 Feed the OTSG with the Auxiliary feed water pump P-318 sr P-319 to maintain 85% to 95% on the operate range, per A.51, Section 7.4.
NQTE: If the period of plant shutdown is sufficiently long, losses to ambient may be adequate to allow reactor coolant temperature to stabilize below 212' F without forced RCS circulation or OTSG steaming: natural circulation is an acceptable means of decay heat removal. However, the normal natural circulation parameters, as discussed in B.4, Section 6.0, will not be continuously evident. At low decay heat loads, natural circulation will be intermittent as
(' sufficient driving force may take hours or days to develop, depending on decay heat. Development of the driving force will be indicated by a divergence of incore thermocouples and cold leg temperatures, as incores heat up. When the AT becomes sufficient to push cold water over the RCP lip and into the core, incores and cold leg ~ temperatures will tend to converge (incores decrease). Termination of this
" slug flow" form of cooling will be indicated by the divergence of incore and cold leg temperatures once again,
- as incores heat up.
5* 2.1 .7.10 Verify Natural Circulation per B.4 Section 6 QR per the above guidance.
NOTE: Continuous operation of an RCP will add heat to the core.
2.1 .7.11 1E Incore thermocouples increase to 240*F, THEN Start one
- RCP for continuous operation.
2.1 .7.12 Stabilize RCS temperature as soon as possible; maintain 50*-30' subcooling margin.
Rev. 5 C.12-3 l
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. RESPONSE (Continued) r'h 2.2 Upon loss of DHS flow to the RCS with the RCS open to atmosphere.
CAUTION: Cooling the core in this manner can release reactor coolant to the Reactor Building. As soon as possible after initiation of flow to the RCS, close the Reactor Building equipment hatch, close personnel access hatch, shutdown the purge supply and exhaust fans and then close the purge inlet / outlet valves.
2.2 .1 Sound the Reactor Building Evacuation alarm and advise Health Physics Division of imminent RCS spill in the Reactor Building.
2.2 .2 Close HV-20001 or HV-20002 (RCS to DHS isolation valves). This will prevent back flow of water in subsequent steps.
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2.2- .3 Jog open SFV-25003 or SFV-25004 (utilize the BWST supply to the in service DHS).
CAUTION: Jog open SFV-25003/25004 and monitor RCS level increase (pressurizer and/or reactor vessel level indicator).
2.3 Trend the BWST level (LOO 2) and throttle SFV-25003/25004 to establish a flow rate of at least 120 gpm (approximately 0.02 feet per minute indicated decrease in BWST Level Trend). 'This will require periodic
-l adjustments as the RCS fills.
S?* 2.4 Notify the Nuclear Ops. Manager and Operation: Unit Superintendent as soon as possible.
2.5 When flooding the RCS as described in step 2.2.1 .4, the following
" overflow" point are reached:
2.5 .1 81 inches (LT-21406) and 33 inches (pressurizer) at RCP seals.
I 2.5 .2 84 inches (LT-21406) and 36 inches (pressurizer at reactor vessel flange.
2.5 .3 About 320 inches (pressurizer) at CRD vent nozzles.
l 2.6 Fill the RCS to overflow if applicable and proceed with step 2.2.12, -OR 2.7 Fill the RCS to 310 inches on the pressurizer level indication.
NOTE: If the RCS can be filled to 310 inches on the pressurizer level indication and control rod vent nozzles can be closed, aN2 bubble can be established in accordance with procedures to provide natural convection core cooling. This l would require Reactor Building entry for suitable valve lineups. If the control rod vent nozzles cannot be secured, l ' the RCS will require periodic drain and refill to reduce t
steaming off of the reactor coolant.
Rev. 5 l C.12-4 l
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4 RESPONSE (Continued)
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2.2 .8 Isolate the Reactor Building normal sump drains to prevent loss of water to the Auxiliary Building.
2.2 .9 Attempt to restart the alternate DHS in accordance with procedures, i.e., fill, vent system lineups, etc.
2.2 .10 Drain and fill of the RCS to the Reactor Building Emergency Sump is established by: .
2.2 .10.1 Isolate the BWST from the DHS suction header.
2.2 .10.2 Open HV-20001 and HV-20002.
2.2 .10.3 Throttle HV-20003 to about 120 to 160 gpm and lower RCS indicated pressurizer level to about 40 inches.
2.2 .10.4 Reclose HV-20001, HV-20002, and HV-20003.
2.2 .10.5 Establish flow from the BWST to the RCS by repeating steps 2.2.1 - 4. -
2.2 .10.6 Restore BWST level using step 2.2.12 as necessary.
2.2 .11 As required, establish a lineup with a CBS pump to take a suction
(. on the Reactor Building Emergency sump and discharge to the BWST.
Run this CBS pump as necessary to insure adequate BWST gravity flow to the reactor.
CAUTION: Do not run the CBS pump (s) unless the Emergency Sump level alarm indicates at least one foot of water in the sump.
2.2 .12 Alternate method for cooling RCS.
NOTE: Flow path is gravity flow to the RCS via "A" Decay Heat Pump, return to BWST by operating "B" CB spray pump.
Alternate flowpath "B" Decay Heat Pump, "A" CBS is in parenthesis.
2.2 .12.1 Verify 120 gpm gravity flow to RCS through A(B) DHR system.
Isolate the BWST from the B(A) DHR suction header, close SFV-25004 (SFV-25003).
2.2 .12.2 Line up DHR drop line to the B(A) DHR suction header, open HV-20006 (HV-20005).
Rev. 5 C.12-5
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l RESPONSE (Continued)
.i . l 2.2 .12.3 Isolate the DHR drop line from the A(B) DHR suction header, close HV-20005 (HV-20006).
2.2 .12.4 Open HV-20001 and HV-20002. l 2.2 .12.5 Line up B(A) CBS recirculation to BWST, open CBS-046 (CBS-045).
2.2 .12.6 Verify spray add tank suction valves closed SFV-29015, SFV-29016.
2.2 .12.7 Start the B(A) spray pump.
2.2 .12.8 Throttle open CBS'BWST recirculation valve CBS 046 (045) to match RCS gravity fill rate. Maintain RCS level just below the over- flow point of step 2.2.6.
2.2 .12.9 Open SFV-26039 (SFV-23040) DHR cooler outlet valve and close HV-26037 (HV-26038) DHR cooler bypass valve to direct all flow through the cooler.
2.2 .12.10 Monitor incore T/Cs and DHR A(B) suction temperature.
Increase gravity flow and CBS pump return to BWST flow as necessary to maintain constant RCS temperature.
- 7. Recirculation temperature to the BWST should be minimized (<
,, 140*F) if possible to reduce BWST heati.ng.
2.3 High Flow Conditions l
2.3 .1 Throttle the operating DH loop A(B) cooler discharge or cooler bypass valves to reduce flow.
2.4 Reactor Coolant Temperature Rise 2.4 .1 Check decay heat discharge temperatures, adjust cooler bypass valves HV-26037 or HV-26038 as required.
2.4 .2 Check NSCW pump (s) running and pump discharge pressure and flow through cooler.
2.4 .3 Check NSRW pump (s) running and pump discharge pressure and flow through cooler.
2.4 .4 Check spray pond bypass valves.
2.4 .5 If high temperature still exists, place second decay heat system loop in service per OP A.8.
t Rev. 5 C.12-6 i
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.,6 3 RESPONSE. (Continued) 2.5 Decreasing pressurizer, fuel transfer canal or reactor vessel level.
2.5 .1 If pressurizer level, fuel transfer canal or reactor vessel is decreasing, stop the operating decay heat pump, then close HV-20001.
2.5 .2 Terminate any operation that may be bleeding water f rom RCS or decay heat system.
2.5 .3 Check R8 sump and decay heat pump room sump levels to determine if level decrease is due to leakage.
2.5 .4 Isolate DH loop leak or place alternate OH loop in service.
2.5 .5 If fuel transfer canal leakage continues, terminate fuel handling operations.
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Rev. 5 C.12-7 l
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SACRAMENTO MUNICIPAL UTILITY DISTRICT kr2 OFFICE MEMORANDUM i
TO: Bob Wichert DATE: February 17, 1987 TS87-047 FROM: Rob McAndrew
SUBJECT:
RCS BORON CONCENTRATION DURING DHS OUTAGE The purpose of this letter is to address the concern that the boron '
concentration in the core may be reduced by slugs of low boron water.
In a discussion with Steve Redeker I learned that the NRC expressed a concern that pockets of water with low boron concentration would be formed and subsequently be swept into the reactor core resulting in an uncontrolled reactivity change. I have investigated this concern and conclude there is no identified mechanism for developing pockets of' low boron water in the RCS for the configuration Rancho Seco will be in during the DHS outage. The basis for this conclusion is:
Boron is in solution in water at temperature of 80-280*F and concentrations between 1800-2200 ppm. No matter how long the solution remains stagnant, the boron will not come out of solution. The fact that boron is in solution and is not a suspension prevents settling out of boron.
The temperature of the RCS and makeup. water is capable of dissolving approximately 10,000 ppm boron at 80*F. This is the lowest temperature expected during the DHS outage.
The highest boron concentration water which will be added to the RCS during the DHS outage is 2200 ppm. Thus there is considerable margin for boron solution capability.
Even if a mechanism for the formation of low boron con-centration water could be presumed, these pockets would be mixed with the RCS during natural circulation flow. The decay heat from the irradiated fuel has been calculated to cause density changes in the RCS water great enough to cause natural circulation. This is based on an evaluation l
of the TMI-2 data with respect to the Rancho Seco config-uration. This will have the effect of circulating RCS coolant between the steam generators, the reactor vessel, and connecting piping. Such action will cause mixing to occur sufficient to preclude pockets of low boron concen-tration water from forming, even if a mechanism could be postulated for initiating the pockets.
i
A 5:b '.icher: -E- February 17, 1987 -
here are two mechanis .s which ::ulc result in boron pockets or stratification to occu , These tre:
Adding a solution of very high boren concentration to that with a low baron c:ncentration or vice versa
~followed by allowing the -ixture to remain absolutely stagnant.
Lowering the temperature :f the water to below the solubility limit of the boren.
Neither of these conditions will be allowed to occur during the DHS outage. In fact quite the opposite is to be expected. The natural circulation expected to occur iil actively promote mixing of any water which could be at a different boron concentration that the rest of the system. Also, the temperature of the RCS will not be allowed to fall to the boron solubility limit for boron.
To further assist with the licensing evaluation of the DHS outage, the following discussion of reacti .ity additions to the core is provided.
The major factor affecting core reactivity during the DHS outage will be.the_ core temperature change. To evaluate the effect of the tem-perature change on core reactivity entails evaluating the net effect moderator reactivity changes, #uel reactivity changes, burnable poison reactivity changes, control rod reactivity changes, and boron reactivity changes. Fortunately, the fuel . cycle designer has performed this analysis.
The fuel cycle designer's calculations, results of which have been placed in OP B.6, indicate the boron required to maintain cold shutdown (All Rods In, Highest Worth Stuck Rod Out,1% delta k/k shutdown) is 1300 ppm, for our current burnup. This calculation was done considering the core to be at 68*F. The same calculation was done for hot shutdown.
HSD. The boron concentration, required to maintain HSD (All Rods In, Highest Worth Rod Stuck Out,'l' delta k/k shutdown) is 720 ppm. Note the boron required at cold conditions is higher than for hot conditions.
This means that if the RCS temoerature increases above 68'F, the required shutdown will be maintained as long as the cold shutdown boron concentra-l tion is maintained in the RCS. Since the RCS boron concentration during the DHS outage will be maintained above 1800 ppm, the required shutdown will be assured for temperature range of 68-532*F.
cc: Jim Field Rob Roehler Masa Nakao Greg Smith Roger Powers Eric Ycchheim
, Steve Redeker 2E.5CO l , ['\-psiw % i w /
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&./ 1 Nchw Pm Divien it a McDermott comcany j315 Olc Fe st Read February 23, 1987 ^ '$ ~'
$$"3yyoco SMUD-87-084 Mr. Greg.Cranston, Manager ~
Nuclear Engineering Department Sacramento Municipal Utility District Rancho Seco Nuclear Generating Station 14440 Twin Cities Road
'Harald, CA 95638
- Attention: Mr. Steve Redeker
Subject:
Rancho DHRS Seco RCS Response Following Termination of k
Reference:
Rancho Seco Nuclear Generating Station, Unit 1 Master Services Contract dated January 1, 198'6 SMUD Contract A977-B&W contract 582-7504 Task 765, Subtask 8 - Incidental Engineering Services for the operations Department
Enclosure:
.B&W ' Document 32-1167923-00, " Rancho Seco RCS Response Following Termination of DHRS" Gentlemen: -
Enclosed following istermination B&W's calculation of of the Rancho Seco RCS response the DHRS. This calculation was requested by deliverable product for the referenced task. the District's Mr. Greg Smith and constitutes the If you have any questions, please contact me at (804) 385-2308 in Lynchburg.
Very truly yours, f
. . Burke nager of Contract Engineering Nuclear Engineering Services i
rRS/rlb cc: G. Smith J.T. Janis D. Mixa l R. Crowley R. Roehler I
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