ML18092A762
| ML18092A762 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/30/1985 |
| From: | Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18092A761 | List: |
| References | |
| NUDOCS 8509090196 | |
| Download: ML18092A762 (28) | |
Text
REAC'IDR COOLANT SYSTEM COLD SHlITOOWN LIMITING CONDITION FOR OPERATION 3.4.1.4
'IWo # residual heat removal loops shall be OPERABLE* and at least one RHR loop shall be in operation.**
APPLICABILITY:
MODE 5.**
ACTION:
- a.
With less than the above required loops OPERABLE, immediately initiate cnrrective action to return the required loops to OPERABLE status as soon as possible.
- b.
With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate cnrrective action to return the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4 At least one residual heat removal loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Qle RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation.
Additionally, the following may be substituted for one residual heat removal loop:
- 1.
Four filled reactor (X)Olant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), or
- 2.
All equiµnent listed on Table 3.4-3.
A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water volume is less than 1650 cubic feet (equivalent to approximately 92% of level), or 2).the secondary water temperature of each steam generator is less than 50°F above each of the RCS cold leg temperatures.
The normal or emergency power source may be inoperable.
- The residual heat removal pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.
SALEM - UNIT 1
(
a509090t96 a;g~g72 PDR ADOCK'O. PDR p
3/4 4-3b
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REACTOR COOLANT SYSTEM TABLE 3.4-3 MINIMUM EQUIPMENT RE QUIRED FOR DECAY HEAT REMOVAL 0
Two Residual Heat Rermval Pumps and Heat Exchangers 0
Two Component Cooling Pumps 0
Two Service Water Pumps 0
Two *** ECCS pumps capable of supporting the Make-up/Boil-off rrethod of heat rerroval
. A combination of any two ECCS pumps shal 1 be avail able for operation *
. One of the available pumps may be tagged out of service in order to comply with the requfrements of Section 3.5.3 Footnote (#) and 4.5.3.2.
Provided that the pump can be. made operable within 15 minutes.
SALEM - UNIT 1 3/4 4-3c
REFUELING OPERATIONS LOW Wl\\TER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Rerroval (RHR) loops# shall be OPERABLE.*
APPLICABILITY:
MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.
ACTION:
- a.
With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b.
The provisions of Specification 3.0.3 are to applicable.
SURVEILLANCE REQUIREMENTS 4.9.8.2 The required Residual Heat Rerroval loops shall be determined OPERABLE per Specification 4.0.5.
- The normal or emergency power source may be inoperable for each RHR loop.
- Except as specified in Section 3.4.1.4 footnote (#).
SALEM - UNIT 1 3/4 9-8a
REACTOR COOLANT SYSTEM COLD SHUTDO YIN LIMITING CONDITION FOR OPERATION 3.4.1.4 Two# residual heat rerroval loops shall be OPERABLE* and at least one RHR loop shall be in operation.**
APPLICABILITY:
MODE 5.**
ACTION:
- a.
With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.
- b.
With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REQJIREMENTS 4.4.1.4 At least one residual heat rerroval loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
One RHR loop may be inoperable for up to two hours for surveillance testing, provided the other RHR loop is OPERABLE and in operation.
Additionally, the following may be substituted for one residual heat removal loop:
- 1.
Four filled reactor coolant loops, with at least two steam generators with their secondary side water levels greater than or equal to 5% (narrow range), or
- 2.
All equipment listed on Table 3.4-3.
A reactor coolant pump shall not be started with one or more of the RCS cold leg temperatures less than or equal to 312°F unless 1) the pressurizer water voTurre is less than 1650 cubic feet (equivalent to approximately 92% of level), or 2) the secondary water temperature 6f each steam generator is less than 50°F above each of the RCS cold leg temperatures.
The normal or emergency power source may be inoperable.
- The residual heat rerroval pumps may be de-energized for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> provided 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10°F below saturation temperature.
SALEM -
UN IT 2
-3/4 4-4a
\\
REACTOR COOLANT SYSTEM TABLE 3.4-3 MINIMUM EQJIPMENT REQJIRED FOR DECAY HEAT REMOVAL 0
Two Residual Heat ReJOOval Pumps and Heat Excha_ngers 0
Two Component Coo 1 i ng Pumps 0
Two Service Water Pumps 0
Two *** ECCS pumps capable of supporting the Make-up/Boil-off rrethod of heat reJOOval A combination of any two ECCS pumps shall be. available for operation.
One of the available pumps may be tagged out of service in.order to comply with the requirements of Section 3.5.3 Footnote(#) and 4.5.3.2.
Provided that the pump can be made operable within 15 minutes.
SALEM - UNIL 2 3/4 4-4b
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REFUELING OPERATIONS LOW -Wl\\TER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent Residual Heat Rerroval (RHR) loops# shall be OPERABLE.*
APPLICABILITY:
MODE 6 when the water level above the top of the reactor pressure vessel flange is 1 ess than 23 feet.
ACTION:
- a.
With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b.
The provisions of Specification 3.0.3 are to applicable.
SURVEILLANCE RE~IREMENTS 4.9.8.2 The required Residual Heat Removal loops shall be determined OPERABLE per Specification 4.0.5.
- The normal or emergency power source may be inoperable* for each RHR loop.
- Except as specified in Section 3.4.1.4 footnote (#).
SALEM - UNIT 2 3/4 9-9
SAFETY EVALUATION FOR OPERATION OF BOTH SALEM PLANTS IN MODES 5 AND 6 WITH ONE SERVICE LOOP OUT OF SERVICE Public Service Electric & Gas Co.
August, 1985
TABLE OF CONTENTS
- 1. 0 INTRODUCTION *....
.. 1 2.0 DISCUSSION OF PROPOSED TECHNICAL SPECIFICATION CHANGE *. 1
3.0 BACKGROUND
ON SERVICE WATER SYSTEM.
- 2
- 4. 0 SAFETY JUSTIFICATION...... *
. 3 5.0 4.1 FSAR Accident Analysis Impact. *.. * *.
- . 3 4.2 High Reliability of Intact RHR Equipment...*. 4 4.3 Diverse Means of Decay Heat Removal *
...** 5 4.4 4.5 4.3.l 4.3.2 4.3.3 Core Uncovery Time... *.
Short Term Decay Heat Removal Long Term Decay Heat Removal
- Non-DHR Service Water Loads Compensatory Actions.
4.6 Overall Improvement in Safety
SUMMARY
AND CONCLUSION
- 6
. 6
- 8
.... 10
. 12
..... 14
.... 14
6.0 REFERENCES
AND FIGURES 16 17 18 19 FIGURE 1 -
SALEM REFUELING EVOLUTIONS. *.
- FIGURE 2 -
SALEM RESIDUAL HEAT REMOVAL SYSTEM FIGURE 3 -
SALEM COMPONENT COOLING SYSTEM FIGURE 4 -
SALEM SERVICE WATER SYSTEM * *
- *. 20
-ii-
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1.0 INTRODUCTION
The service water system at the Salem Nuclear plants utilizes water from the Delaware River as its heat sink.
This water source is typically brackish and has a relatively high silt content.
In order to ensure high reliability and integrity of the service water system, PSE&G has decided to conduct detailed inspections of the service water system, for both Units 1 and 2, during each refueling outage.
Each Unit has two service water headers.
In order to perform these inspections it is necessary to take each service water header out of service, during the refueling outages, for an extended time period while the plant is in modes 5 and 6.
The current Salem technical specifications governing these modes of operations do not allow a service water header to be out of service for a long enough period of time to conduct all desired inspections.
The purpose of this safety evaluation is to demonstrate that the proposed changes to the Salem technical specifications, allowing one service water header to be out of service in modes 5 and 6, do not result in any significant degradation of plant safety during this relatively short time period.
The end result of the change wilL be an increase in the reliability and integrity of the service water system, with an associated improvement in safety.
2.0 DISCUSSION OF PROPOSED TECHNICAL SPECIFICATION CHANGE The current Salem technical specifications require residual heat removal (RHR) loops to be available in modes 5 and 6 as follows:
Mode 5 Mode 6 Two RHR loops are required to be operable if no steam generators are available.
Two RHR loops are required to be operable if the water level in the refueling cavity is less than 23 feet above the reactor vessel flange.
Since service water supplies the ultimate heat sink for the RHR system, the current requirement to have two RHR loops operable requires that both service water headers be available.
Figure 1 shows how water inventory in the reactor ves~el changes with time during the anticipated 60 day refueling outages.
In order to conduct the desired inspection on the service water system, one service water loop will be out of service for a 32 day period (16 days for each header).
As shown on Figure 1, there are two time periods (about ten days duration each) when a service water loop will be out of service, which is not allowable with current technical specifications.
The first period starts when the draining is initiated to bring the water level down to the centerline of the nozzle and ends when the water level is raised above 23 feet for refueling evolutions.
During this period, necessary steam generator eddy current inspections are performed, reactor coolant pump maintenance is conducted, and the reactor vessel head is removed.
- Thus, during the first period, the reactor coolant system is open via either the open steam generator manway or open reactor vessel.
The plant transitions from mode 5 to mode 6 during this period when the reactor vessel head studs are detensioned for head removal.
The second period starts as the water level passes through the 23 foot mark on its way to the reactor vessel flange in preparation for reinstallation of the reactor vessel head, following refueling.
The period ends when the reactor coolant system has been filled and vented.
During this period, the reactor vessel head and associated equipment is installed, the primary coolant system is closed, and the necessary valve lineups are performed to allow the primary coolant system to be filled and vented.
The plant transi-tions from mode 6 to mode 5 during this period when the reactor head studs are tensioned.
Of the two time periods discussed above, the first period is the most critical since the primary coolant inventory is at the minimum level and the decay heat is high.
During the second period, the decay heat is significantly reduced since the core now contains fresh fuel, the time after shutdown is longer, and the primary water inventory is greater (water level at the vessel flange).
3.0 BACKGROUND
ON THE SERVICE WATER SYSTEM During modes 5 and 6 the service water system serves as the normal heat sink for loads associated with decay heat removal and is the stand by heat sink for certain emergency equipment heat loads.
The equipment serviced by the service water system under these conditions is as follows:
Normal Operating Heat Loads o
Component cooling water
--RHR heat exchanger
--RHR pump seal heat exchanger
--Spent fuel pit heat exchanger o
Chiller condenser o
CCW and RHR pumproom coolers o
Containment fan cooling units Emergency Equipment Heat Loads o
Diesel generator (lube oil and jacket water coolers) o Safety injection pump (lube oil and seal coolers) o Charging pump (lube oil and seal coolers) o Room coolers (safety injection, charging, containment spray and auxiliary feedwater pumps)
In the proposed configuration, with one service water loop out for maintenance, the failure of the remaining service water system would result in the complete loss of service water.
The impact of this unlikely (given the compensatory actions and the redundancy in the operating service water loop) complete loss of service water to the normal and emergency loads is addressed in the following safety justification.
4.0 SAFETY JUSTIFICATION (SIGNIFICANT HAZARDS EVALUATION) 4.1 FSAR Accident Analysis Impact The majority of the accidents discussed in Chapter 15 of the Salem FSAR are not applicable when in modes 5 and 6 with the reactor coolant system depressurized.
The reactivity insertion events involving the control rods are not applicable since all control and shutdown banks will be inserted into the core and de-energized during the proposed mode of operation.
The loss of flow, loss of feedwater, feedwater system malfunction, excessive load increase, depressurization, spurious safety injection and turbine generator accidents are not.applicable since the equipment assumed to fail will not be in operation in modes 5 and 6.
The pipe rupture events are not applicable since both the primary and secondary system will be depressurized.
The only FSAR accidents that could occur in modes 5 and 6 are loss of offsite power and boron dilution.
The concerns with loss of offsite power, addressed in the FSAR, relate to the associated loss of primary and secondary flow which are not a concern in modes 5 and
- 6.
A loss of offsite power could have an effect on decay heat removal capability.
This issue is addressed in Section 4.4 of this safety evaluation.
Relative to the boron dilution accident, the proposed configuration, with one service water header out of service, has no impact on either the probability of occurrence or the capability to mitigate the conse-quences.
The same emergency boration paths, assumed in the current FSAR, will still be available.
Thus, the proposed technical specification change has no impact on the Salem FSAR accident analysis.
4.2 High Reliability of Intact RHR Equipment A simplified flow diagram of the RHR, CCW, and service water systems is given in Figures 2, 3, and 4, respec-tively.
As seen on the flow diagrams, the Salem plant design normally includes redundant loops for the RHR, CCW, and service water systems.
Only one loop of each system is required to maintain decay heat removal.
In the configuration proposed herein, even though one service water loop will be out for maintenance, both loops of residual heat removal and component cooling water (CCW) will be kept available, consistent with the requirements of the current Salem technical specif ica-tions.
A minimum of two RHR, two CCW, and two service water pumps, powered from two different vital busses, will be kept available.
Thus, normal redundancy is retained down to the service water system.
Only one component cooling water heat exchanger will be available since orily one service water loop is avail-able.
The CCW heat exchangers for both Units 1 and 2 have recently been retubed and therefore have a very high reliability.
Even if a tube leak did occur, decay heat removal capability will be maintained since the service water system pressure is higher that the CCW system so that no loss in CCW inventory would occur.
The only remaining active components that could poten-tially defeat RHR heat removal are the two series RHR hot leg suction valves and a limited set of valves in the service water system flow path to and from the CCW heat exchangers.
A review of historical failure rate data, based on the current Nuclear Plant Reliability Data System (NPRDS) database, was performed to deter-mine the reliability of these remaining active failure points.
A reliability analysis of the system configur-ation evaluating the combined reliability of all the single point active valves for the time period of concern was conducted.
The results of this evaluation show that the combined reliability of these active valves is greater than 0.99.
As discussed in Section 4.5 below, additional actions will be taken to effectively eliminate any possibility of these single point valves from failing and def eating RHR heat removal.
That is, all the single point valves will either be disabled in position (power removed for motor operated valves) or locked in position (manual valves physically locked), with the exception of one air operated temperature control valve on the outlet from th~ CCW heat exchanger.
This includes the two RHR hot leg suction valves which will be disabled in the open position when the reactor coolant system is depressurized and open and there is no need for over-pressure protection of the RHR system.
This represents an improvement in safety even over the normal residual heat removal methods when both service loops are available.
With the compensatory actions to be taken to ensure single failure point valves are maintained in position, all potential active failures have been eliminated with the exception of one air operated temperature control valve.
This particular valve fails open on loss of air which is the safe position.
This valve has never failed in the closed position, which can only occur due to a spurious signal to the temperature controller, and is therefore considered to have a very high relia-bility.
At most, this failure would result in a brief degradation in residual heat removal capacity, but would not def eat the RHR system for any significant amount of time.
The RHR temperature, monitored by the operator, would increase and alert the operator to take prompt action to reopen the valve.
Therefore, there are no credible single active failures that can defeat residual heat removal.
Thus, with one service water header out of service, the intent of the technical specifications as defined in the bases section of the tech. specs. (to have a single failure proof RHR system) is met with the proposed system configuration.
4.3 Diverse Means of Decay Heat Removal Decay heat removal with one service water loop out for maintenance is very reliable as discussed in the previous section.
This decay heat removal configuration is backed up by additional cooling methods which is consistent with the industry's defense in depth concept.
These additional methods provide time to restore the normal decay heat removal paths or alone can provide for long term decay heat removal.
The available response times to assure adequate measures are taken to maintain core cooling in modes 5 and 6 are significantly longer than those available during normal operation.
This is demonstrated in the evaluation of core uncovery time, Section 4.3.1.
The heat capacity of the refueling water storage tank and the spent fuel pools can be used as a methods of increasing the time available to restore the normal residual heat removal systems.
This method of short term decay* heat removal is discussed in Section 4.3.2.
An alternative for long term decay heat removal is available through the initiation of a simplified make-up and boil-off process.
This process by itself is adequate to satisfy long term decay heat removal requirements.
This method of decay heat removal is discussed in Section 4.3.3.
4.3.1 Core Uncovery Time Calculations have been performed to determine the amount of time available, assuming no operator inter-vention, before the available reactor vessel water inventory boils off and the core begins to uncover.
Core uncovery times were based on the conservative decay heat formulation given in Reference 1 and the water volumes given in Table 5.2.3 of Reference 2.
During the first period, when the water level is at the nozzle centerline, the time to core uncovery varies from one and a half (1 1/2) hours at the beginning of the period to two and a half (2 1/2) hours at the end of the period.
Based on a more realistic decay heat generation, these core uncovery times are actually expected to be on the order of two to three hours, from the beginning to the end of the time period, respec-tively.
During the second time period, the core uncovery time would be much greater (>10 hours) since the water level is up at the reactor vessel flange and the decay heat at this time is lower.
If normal residual heat removal is lost during either of these time periods, adequate instrumentation exists to quickly alert the operator to the problem and suff i-cient time exists to either restore RHR heat removal or initiate the alternate heat removal means discussed below.
4.3.2 Short Term Decay Heat Removal An alternate method of short term (several hours) decay heat removal is available to provide operators and maintenance personnel time to restore normal RHR cooling prior to the initiation of make-up and boiloff.
This method makes use of the refueling water storage tank and spent fuel pit water inventories as heat sinks.
To utilize the RWST heat sink, cold water would be drawn from the RWST by one containment spray pump which would be aligned to discharge into the cold legs of the reactor coolant*system.
This water would remove core decay heat, heating up to approximately 150°F, and be returned to the RWST by an RHR pump taking suction off the hot leg.
The d5sign limit for the refueling water storage tank (120 F) would limit decay heat removal in this mode.
Either in parallel or series with the RWST heat removal scheme, the spent fuel pits of both Units 1 and 2 will be utilized as heat sinks.
Component cooling water would remove heat from the RHR heat exchanger in the same manner as during normal RHR system operation.
Component cooling water would then be circulated through the spent fuel pit heat exchanger transferring its heat to the spent fuel pool.
After the fuel pool of the faulted Unit (the one without service water) has reached the allowable temperature, an existing cross-connect between Unit 1 and 2 will be used to circulate the intact Unit's fuel pool water through the spent fuel pool heat exchanger of the faulted Unit.
The fuel pools will only be permitted to heat up to 100°F.
This will allow for several days with no cooling of the fuel pools before boiling occurs.
The combination of the two methods of heat removal discussed above can maintain normal decay heat removal for a period ranging from two to nine hours, during the critical period when the water level is at the nozzle centerline.
These times are based on the decay heat curve given in Reference 1 without the 10% uncertainty factor applied.
The time available depends on the length of time after shutdown that service water is lost and the initial temperature of the RWST and the fuel pools.
The temperatures of the fuel pools is dependent on service water temperature (Delaware River temperature).
The two hour time period would apply if both the RWST and the spent fuel pit are initially at a maximum temperature of 95°F and the intact service water system is lost four days after shutdown (first day when the reactor vessel water level is at the nozzle centerline).
If the initial temperature of the RWST and the Delaware River is assumed to be 80°F, the minimum time available four days after shutdown would be five hours.
The nine hour heat removal time corresponds to the end of the ten day period, when the water level is at the nozzle centerline.
For the time periods when refueling outages are scheduled (spring or fall) the average monthly temperature of the spent fuel
- ~t
- q~~~ ~~e 8
~~~~ can typically be maintained less than For the first ten day period of concern with the vessel water level at the nozzle centerline, the water level in the reactor vessel can also be increased in order to maximize the water inventory above the core.
As discussed earlier, the reactor coolant system will be open during this ten day period via either an open steam generator manway or the reactor vessel head being removed.
If a steam generator is open, it needs to be closed up prior to filling the system.
This would be accomplished while the short term heat removal process is being utilized.
The manway covers or a temporary blank flange can be installed in a relatively short time (<2 hours).
Then, the vessel can be filled up to at least the reactor vessel flange.
This would increase the core uncovery time to greater than six hours if no heat removal is available (e.g., after the short term heat removal capacity is exhausted) and would allow more time to initiate the long term heat removal process, if required.
Based on a review of PWR operating experience, the majority of loss of decay heat removal events that have occurred have only lasted for a relatively short duration.
Only a few events have lasted longer than one hour and no event has lasted more than three hours.
Thus, if a loss of RHR heat removal should occur, the alternate means of decay heat removal discussed above should provid~ adequate time to restore normal decay heat removal via the RHR system.
It should be noted that this short term decay heat removal process is not required to sustain a safe shutdown condition.
Rather, the short term heat removal scheme will be employed mainly for economic considerations to avoid having the decontamination and cleanup problems associated with the make-up and boil-off process.
This long term alternate decay heat remove scheme, discussed in the next subsection, is adequate by itself to maintain the required decay heat removal.
. 4. 3. 3 Long Term Decay Heat Removal In the unlikely event that an alternate means of decay heat removal will be needed for a time period longer than those discussed above, a make-up and boil-off process can be used.
Water from the RWST can be fed into the cold legs by any one of the ECCS pumps (safety injection, charging, residual heat removal, or containment spray pumps) and allowed to boil off through either the open manway in a steam generator or up through an open reactor vessel.
Any one of the ECCS pumps can be used for this process as long as appropriate actions are taken (discussed in Section 4.4) to ensure the long term operability of the pumps.
Based on the conservative decay heat curve generated from Reference 1, approximately 80 gallons per minute would be required to maintain the water level at the centerline of the nozzle at the beginning of the critical time period (first ten day period for which technical specification relief is required).
With this flow rate, decay heat can be removed for greater than three days with the water inventory in the RWST alone.
This is equivalent to the heat sink available when the refueling cavity is filled to 23 feet (currently technical specifications require only one RHR loop to be operable during this period).
This time can be extended even further through use of makeup from the demineralized water* tanks to the refueling water storage tank.
For the second ten day time period with the water level up to the reactor vessel flange, this method could also be used with boil-off through either an open vessel or through the Pressurizer power operated relief valves.
In order to ensure that this method of heat removal is available during the critical time periods, a minimum of two safety grade pumps (any combination of safety injection and charging pumps) will be maintained available (won't be taken out for maintenance).
If the make-up and boil-off method of heat removal should be required, the containment pressure then becomes a potential concern.
If boil-off occurs at the rate discussed above (-80 GPM), it is estimated, based on the conservative decay heat curve generated from Reference 1, that it will take at least ten hours to reach containment design pressure.
In order to ensure that containment pressure is maintained well below design pressure, the containment would be periodically vented.
Detailed dose analyses have been performed to determine the offsite consequences of venting.
The methodology used for the dose calculations was consistent with that given in Reference 3 (R.G. 1.109).
The dose rate conversion factors (DRCF) were based on site specific characteristics for a mixture of noble gas and iodine and a minimum of 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> decay.
The average X/Q values were obtained from Reference 2 (Table 2.3).
Based on the coolant activity expected, the offsite doses are below both the 10CFRlOO limits and the normal operation Appendix I limits.
Based on the maximum allowable technical specification coolant activity and the maximum purge rate, the offsite doses are signifi-cantly below 10CFRlOO limits.
Based on both the low historical coolant activity at Salem and coolant clean up activities, via the chemical and volume control system, prior to lowering the water level to the nozzle centerline, the coolant activity will be significantly below technical specification limits.
The make-up and boil-off method of heat removal described above can also be used if power is unavailable.
This is accomplished by gravity feed from the RWST to the reactor vessel.
The water level in the RWST is at an elevation approximately 45 feet above the reactor vessel centerline, providing an adequate driving head to supply water to the core.
The f lowrate would be controlled locally by throttling one of the valves downstream of the RWST outside containment.
4.4. Non-DHR Service Water Loads As discussed previously, there are several heat loads serviced by the service water system which are not directly associated with core decay heat removal.
If a total loss of service water should occur, these heat loads can either be supplied by alternate heat removal methods or are not required for an extended period of time.
The heat loads supplied by service water, either directly or indirectly during normal operation in modes 5 and 6, include the chiller condensers, the RHR and ccw pump room coolers, the containment fan cooling units, the RHR pump seal heat exchangers and the spent fuel pit heat exchanger.
The chiller condensers supply cooling for the control room, protection cabinets, relay room and plant computer.
Normal supply and exhaust ventilation fans will still be available to supply ventilation with outside air, if service water is lost.
Based on normal operation design heat loads ang a conservative initial outside air temperature of 95 F, the peak temperature reached in these areas is below the design limit for all safety related equip-ment.
The plant computer will exceed its design temperature; but, it is not needed to sustain a safe shutdown condition.
If service water is lost for an extended period of time, a currently existing cross-connect between Salem Units 1 and 2 can be used to supply all the necessary chilled water.
If service water is lost, the CCW and RHR pumps would be needed for the short term heat removal scheme discussed in section 4.3.2 above.
Both of these pumps can operate within design limits without room coolers for an extended period of time by using portable fans to supply additional ventilation.
The containment fan cooling units are used in modes 5 and 6 to maintain desirable environmental conditions (e.g., low humidity) but are not needed to remove any significant heat loads.
Therefore, the containment fan cooling units are not required during normal operation.
The fan cooling units are not required with the make-up and boil-off process discussed above since containment pressure can be maintained below the design limit with periodic venting.
The RHR pump seal exchangers would only be required for the short time period involved with the short term heat removal scheme discussed in Section 4.3.2 above.
Alternate cooling to this seal heat exchanger can be supplied with a temporary hose connection from the plant demineralized water system.
The discharge from the heat exchanger would be directed to a floor drain in the RHR pump room.
Loss of cooling to the spent fuel pit heat exchangers can be tolerated for a long period of time (several days) before the fuel pool heats up to an unacceptable temperature.
In addition, a currently existing cross-connect between spent fuel pools in Units 1 and 2 can be used for cooling.
The additional heat loads supplied by service water under emergency conditions include lube oil coolers and seal heat exchanger for both charging and safety injection pumps.
The pump lube oil coolers and seal heat exchangers would only be required if the long term make-up and boil-off heat removal scheme discussed in Section 4.3.3 above is required.
Both lube o~l coolers and seal heat exchangers can be supplied with cooling water from the demineralized water system through temporary hose connections.
Service water also provides cooling for the auxiliary feedwater pumps which are not required during the time period of concern.
Service water also provides room cooling for the containment spray pumps which would be required with the short term heat removal scheme, discussed in Section 4.3.2 above.
These pumps can operate for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without room coolers which is greater than the time the pumps would be required.
Service water also provides room cooling for the safety injection and charging pumps which would only be required with the long term make-up and boil-off scheme discussed above.
The safety injection pumps can operate within design limits for an indefinite amount of time without a pump room cooler.
The charging pumps can operate for a long period of time (>12 hours) without room coolers.
Portable fans can also be used to supply any additional ventilation in these pump rooms if desired.
Service water also provides cooling for the diesel generators which would only be required if a loss of service water to the diesel generators occurred coincident with loss of offsite power.
Even if offsite power is lost, it is highly likely that it would be restored in less than an hour, before core uncovery would begin.
Additionally, an existing gas turbine can be used to supply emergency power if required.
The combined probability of loss of service water, loss of offsite power for greater than one hour, and failur~ 6 of the gas turbine to start on demand is so small (<10
)
that this is not considered a credible event.
Core cooling, even with this set of circumstances, can still be accomplished using the make-up and boil-off process with gravity feed from the RWST as discussed in Section 4.3.3.
4.5 Compensatory Actions In order to minimize the likelihood of loss of service water, several compensatory actions will be taken.
As discussed earlier, all single failure point valves will be either disabled or locked open, with the exception of one air operated temperature control valve on the outlet piping of the CCW heat exchangers.
This valve will be monitored closely so that prompt operator action can be taken if the valve should spuriously close.
The operators will closely monitor the critical temper-ature and pressure readings in order to detect any problem that may develop and take appropriate corrective action.
A new procedure will be written to identify the steps that must be taken before entering into the desired configuration (water level at nozzle centerline with one service water loop out for maintenance) and the criteria that must be met.
The procedure will, as a minimum, include the following:
o A listing of valves to be locked open or disabled o
A limit on the allowed coolant activity o
A summary of equipment that must be maintained available (in addition to that already required by Tech. Specs.):
Two RHR, CCW and Service Water pumps One containment spray pump Two ECCS pumps that have the capability to support the make-up and boil-off methods of heat removal The backup gas turbine o
Verify that any temporary hosing, required for the short term and long term backup heat removal, is available in pump rooms.
o Verify that portable fans are easily available to supply ventilation in the pump rooms, if needed.
The currently existing emergency procedures will also be enhanced to specifically identify steps to be taken if a loss of the intact service water loop should occur.
The elements to be added to the procedure will include:
o A plot of core uncovery time versus days after shutdown to give the operator a better understanding of how much time is available.
o A requirement to initiate the short term backup cooling procedure discussed in Section 4.3.2 above if normal RHR heat removal is not restored within a specified time limit.
o A requirement to immediately replace the equipment hatch.
o A listing of the valve lineups and steps to be taken for initiating the short term backup cooling scheme.
o A criteria for switching from the short term backup cooling to the long term backup cooling.
o A listing of the valve lineups and steps to be taken for initiating the long term backup cooling scheme (make-up and boil-off).
o Criteria for venting the containment when in the make-up and boil-off mode.
4.6 Overall Improvement In Safety The proposed technical specification changes will result in an overall improvement in plant safety due to several factors discussed below.
In order to perform maintenance and inspection on the service water system with the current technical speci-fications, each loop must be tagged-in and tagged-out several times during the outage to coincide with current allowed technical specification restrictions.
The large diameter of the service water piping (up to 30 inches) and the location of much of the piping below the river level require significant time and resources to isolate and pump out the lines prior to performing inspection and maintenance. This results in poor utilization of time and manpower resources.
This also increases the chances of operator error when tagging the system in and out.
In addition, due to the long time involved to tag a system in and out and prepare it for inspection, this limits the amount of time avail-able during the outage for inspections on the service water system.
The efficiency associated with one single outage on each service water header will permit the implementation of a comprehensive inspection and maintenance program.
Since the service water system draws water from the Delaware River, which has a high silt content and is brackish in nature, the capability to perform increased inspection and maintenance will result in higher reliability and integrity of the system.
The compensatory actions being taken to disable the two RHR hot leg suction line valves, when the system is depressurized and open, eliminates one of the major contributors to past loss of DHR events (spurious closing of valves).
Combined with the other compensa-tory actions being taken, there are no credible active failures that can defeat normal decay heat removal during the time periods of interest.
Thus, even with one service water loop out for maintenance, the residual heat removal system is single failure proof.
5.0
SUMMARY
AND CONCLUSION The proposed technical specification changes are required to allow having one service water loop out for maintenance for two periods of time in modes 5 and 6, when the steam genera-tors are unavailable and the water level in the refueling cavity is less than 23 feet above the vessel flange.
This change will allow optimization of inspection and maintenance activities during refueling outages and permit more detailed inspections of the service water system.
The following points summarize the safety hazards considerations for this change:
o This change has no significant impact on the Salem FSAR accident analysis.
o High reliability of active components in the operable service water loop.
o Compensatory actions to be taken which include:
elimination of single active failure points; a new procedure to be written for initiating the service water loop maintenance; enhancements to current emergency procedures to address loss of all service water.
o Both short term and long term alternate safety grade decay heat removal methods identified.
This technical basis has been developed to demonstrate that this change will not result in any significant hazards that would degrade plant safety while in modes 5 and 6.
The overall plant safety should be improved due to an increased reliability of the service water system.
Based on this evaluation, we have determined that the No Significant Hazard Consideration criteria of 50.92 have been met.
In particular, we find the following:
o No significant increase in the probability or consequences of accidents previously evaluated exist because of this change.
o No new accident not previously evaluated will be created by operation with this change in effect.
o Margins of safety are improved by implementation of the measures and methods described in this change.
6.0 REFERENCES
- 1.
NRC Branch Technical Position ASB 9-2, "Residual Decay Energy for Light Water Reactors for Long Term Cooling".
- 2.
Salem Generating Station, Units 1 and 2 Updated Final Safety Analysis Report.
- 3.
Regulatory Guide 1.109, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I, Revision 1", October 1977.
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SERVICE WATER NUCLEAR 2 UNIT.
FIGURE 4 SALEM SERVICE WATER SYSTEM
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