ML20136H004
| ML20136H004 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 03/11/1997 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20136H000 | List: |
| References | |
| NUDOCS 9703190028 | |
| Download: ML20136H004 (21) | |
Text
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l UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY j
i I
PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET Nos. 50-282 i
50-306 1
REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 I
LICENSE AMENDMENT REQUEST DATED January 29,1997 i
Amendment of Coolina Water System Emeraency intake Desian Bases Northern States Power Company, a Minnesota corporation, by this letter dated March 11,1997, with Attachments 1 and 2, provides supplemental information in support of the subject license amendment request dated January 29,1997. contains the description of the license conditions and the supporting safety evaluation and significant hazards considerations. Attachment 2 provides the earthquake procedure, AB-3.
This letter and its attachments contain no restricted or other defense information.
NORTHERN STATES POWEP, COMPANY By JogfP. sorense'n' Plant Manager Prairie Island Nuclear Generating Plant On this ll 4 day of % cue L before me a notary public in and for said County, personally appeared, Joel P. Sorensen, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of i.
his knowledge, information, and belief the statements made in it are true and that it is not interposed for delay.
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Supplement 5 to License Amendment Request Dated January 29,1997 Amendment of Coolina Water System Emeraency intake Desian Bases f
Description of the Proposed Changes and the Supporting Safety Evaluation / Significant Hazards Determination Pursuant to 10 CFR Part 50, Sections 50.59 and 50.90, the holders of Operating Licenses DPR-42 and DPR-60 hereby propose the following changes to the licensing bases for the Facility Operating Licenses and Bases for Appendix A, Technical Specification:
NSP requests that the NRC issue the license amendment request dated January 29, 1997 with license conditions discussed below to allow Prairie Island Unit 2 to resume power operations. The proposed license conditions provide interim measures for resolution of the unreviewed safety questions identified in Reference 1and thereby provide continuing assurance that Prairie Island will be operated in a manner which protects the health and safety of the public.
NSP proposes three license conditions as follows-License Condition 1) NSP will provide a licensed operator in the control room on an interim basis for the dedicated purpose of identifying an earthquake. This operator will 4
be in addition to the normal NSP administrative control room staffing requirements and will be provided until License Condition 2 is satisfied.
i License Condition 2) NSP will submit dynamic finite element analyses of the intake canal banks by May 1,1997 for NRC review. Within 30 days of completion of the resolution of NRC questions on the dynamic finite element analyses, NSP will evaluate further required actions and provide a schedule to the NRC for final resolution of the unreviewed safety question. This determination will address whether the time for operator actions can be extended through analyses, physical modifications or some combination. The dedicated operator will be eliminated when the justification based on the analyses is completed or physical modifications are completed.
License Condition 3) Based on the results of License Condition 2, NSP will revise the l
Updated Safety Analysis Report to incorporate the changes into the plant design Page 1 i
bases. These changes will be included in the next scheduled revision of the Updated Safety Analysis Report following completion of License Condition 2 activities.
Reference 1 states that the, "
. NRC determined that taking credit for the non-seismic intake canal and operator actions following an earthquake involved an unreviewed safety question." It is NSP's intent in this submittal to demonstrate that taking credit for operator actions with the use of a dedicated operator, as proposed in License Condition 1, provides a reasonable interim basis for safe operation of Prairie Island.
The Prairie Island normal administrative requirements for control room staffing exceeds the requirements of 10 CFR Part 50,50.54(m).
Resolution of the unreviewed safety questions on a long term basis are dealt with in License Condition 2 and 3. In accordance with License Condition 2, NSP will provide long-term resolution of the unreviewed safety question which takes credit for operator actions within a more relaxed time frame such that use of a dedicated operator is not required. The long-term basis could be seismic evaluation of the intake canal, physical modifications to the intake canal or some combination which assurance that the operators will have sufficient time to address cooling water system loads following a design basis earthquake. License Condition 3 requires incorporating these long-term measures into the plant design basis through revision of the Updated Safety Analysis Report.
BACKGROUND The subject license amendment request was submitted to resolve unreviewed safety questions relating to the Prairie Island cooling water system emergency intake line. A summary discussion of the cooling water system and the issues surrounding the cooling water emergency intake line unreviewed safety questions follows. A more thorough description of the cooling water system, the unreviewed safety questions and NSP's disposition of the unreviewed safety questions can be found in the License Amendment Request Dated January 29,1997, entitled, " Amendment of Cooling Water System Emergency intake Design Bases."
The Prairie Island " Cooling Water System" is similar to the system commonly known in the nuclear industry as the " Service Water System." During the preparation and self assessment activities for the Prairie Island Service Water System Operational Performance Inspection (SWSOPI), an issue was identified by plant personnel concerning the capacity of the cooling water emergency intake line. The emergency intake line pre-operational flow test was reviewed and the results were compared to original design calculations. The results of the pre-operational flow test were less than the design value stated in the Updated Safety Analysis Report (USAR). It was decided Page 2 j
l that the pre-operational flow test should be repeated. The results of this test indicated that there had been some reduction in flow capacity since the pre-operational test. A safety evaluation was prepared to evaluate the effects of emergency intake line flow capacity less than the design value.
The NRC subsequently reviewed the safety evaluation and concluded that the safety l
l evaluation and remedial actions implemented by the Prairie Island staff were acceptable for continued operation of the plant. However, the NRC concluded that taking credit for the non-seismic intake canal and operator actions following an earthquake involved unreviewed safety questions (Reference 1).
The intake canal was evaluated for liquefaction potential and slope stability. The evaluation consisted of additional soil borings, cone penetration tests and pseudostatic evaluation. This evaluation is documented in " Intake Canal Liquefaction Analysis" performed by STS Consultants Ltd. This information was presented to the NRC during meetings on February 24,1997. The NRC identified a need for additional analyses.
The dynamic finite element analyses are in progress.
PURPOSE OF THIS SUPPLEMENT Following a seismic event, the cooling water system flow must be reduced to match the capacity of the emergency cooling water line. The length of time available to reduce the cooling water system flow is dependent on whether the intake canal banks will remain stable or whether they will fail in a manner that will block the plant screenhouse.
Currently, dynamic finite element analyses are being performed. The output of these analyses will be used to further assess the behavior of the intake canal banks during a design basis seismic event.
The design basis earthquake is assumed to occur. There is no reliance on equipment that does not have seismic qualification to mitigate the event. Specifically, loss of offsite power, loss of instrument air, and loss of condensate storage tank inventory is assumed. These assumptions place the maximum demand on the cooling water system. Additionally, a single active failure is assumed. The plant has previously been analyzed for the ability to maintain safe shutdown during a loss of offsite power, assuming a single failure. However, the ability to reduce cooling water demand assuming a single failure is specifically addressed in this submittal.
The probability of a design basis earthquake at Prairie Island is very low. The probability of a design basis earthquake occurring within a specific, short time interval (9 months) is lower yet. The probability of a design basis earthquake occurring within a i
l specific short time interval and a single failure is extremely low. Therefore, the Page 3
l proposed interim operator actions to reduce cooling water system demand provide l
reasonable assurance that the public health and safety will be maintained.
This supplement presents further evaluations which demonstrate that on an interim basis:
the cooling water loads can be managed in the time available using conservative l
assumptions on the available water supply.
potential entrained debris will not negatively impact the operation of the safeguards cooling water system.
the probability that the design basis earthquake and the other assumed failures will occur is low.
Thus, the interim measures proposed are acceptable and reasonable.
SAFETY EVALUATION 1.
Coolina Water Loads Can Be Manaaed in the Time Available I. A.
Time Frame For Operator Action in the interim, operator action is proposed to reduce cooling water system demand in a limited time frame. To establish a clear boundary for calculating available water volumes, an assumption is made that no make up is available from the intake canal into the plant screenhouse. Although this assumption makes calculations more precise, it is very limiting. There are several considerations that indicate that there would be some quantity of make up flow from the intake canal. Specifically:
The STS evaluation is considered to be a best estimate (with conservatisms) evaluation. It is not the conservative calculation needed for a design basis evaluation. However, it provides some assurance that the intake canal banks will not undergo gross failure.
A flow area of 70 ft' through the intake canal will support a flow rate of 31,750 gpm 2
at a water velocity of only 1 ft/sec. This 70 ft flow area is a 94% reduction of the nominal intake canal flow area.
A 75% reduction of the nominal volume of the intake canal will provide 31,750 gpm for 50 minutes with only a 2 ft decrease in the intake canal and plant screenhouse i
level.
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This indicates that with only a small percentage of the intake canal available, the time for operator action could be extended.
l.A.1 Assumptions No inventory for make up is available from the intake canal outside of the Plant Screenhouse.
The traveling screens have no power for motive force and no water for backwash.
Both sluice gates are OPEN.
Two safeguards cooling water pumps are running.
Water level is initially at 674'-6", normal river level.
Water contained in the horizontal portion of the circulating water piping that exits the screenhouse gravity drains back to the Plant Screenhouse.
The total flow demand on the water volume in the Plant Screenhouse is 31,750 gpm. The demand from the cooling water system is calculated to be 29,750 gpm.
An additional 2,000 gpm is from the diesel fire pump.
River level drops to 669' at time = 0, remains there for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
There is a dedicated control room operator, above the current normal administrative minimum, assigned to detect the seismic annunciator and imroediately notify the shift supervisor.
l.A.2. Plant screenhouse and emeraency intake line flow demand The maximum calculated cooling water system flow demand is 29,750 gpm.
Additionally, the diesel fire pump will be available to start and is conservatively assumed to be at its design flow rate of 2,000 gpm. This results in a combined flow demand on the water volume in the Plant Screenhouse of 31,750 gpm. The cooling water system flow demand assumes loss of offsite power and loss of instrument air to produce the maximum cooling water demand. Also, two safeguards cooling water pumps are assumed to be running.
l.A.3. Water volume available it is assumed that there is no make up flow from the intake canal into the Plant Screenhouse. The only water volumes available to supply the cooling water pumps and the fire pump are the Plant Screenhouse and the circulating water l
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piping that runs on the horizontal, outside of the Plant Screenhouse. The emergency intake line is also available to supply the cooling water pumps.
The volume of water available inside of the Plant Screenhouse is 385,875 gallons. This is the free volume of the intake bay plus the safeguards bay. Water level is assumed to be at normal level,674'-6", Only the water volume in the intake bay above elevation 664' was used in this calculation. This level was used to assure an adequate driving head for full flow through the sluice gate.
The circulating water piping is 84" diameter pipe. It runs horizontally out of the Plant ScreenhoJse for 60'-6" before turning upwards. When the circulating water pumps stop due to loss of power, the water in this pipe will gravity drain back to the Plant Screenhouse. This portion of pipe provides an additional 69,850 gallons of water. The circulating water piping is buried in compacted fill, which provides reasonable assurance that the pipe will remain intact.
The total available water volume is 455,725 gallons. Assuming a flow demand of 31,750 gpm, there is 14.4 minutes of water available. However, there are two factors that extend this time frame. The first is operator action to reduce cooling water system flow demand. This is discussed later. The second is the relationship between the emergency intake line and the safeguards bay.
At an initial Plant Screenhouse level of 674'6" and an assumed river elevation of 669'-0", there will be some flow of water from the safeguards bay out through the emergency intake line to the river. The flow rate to the river through the emergency intake line will decrease as Plant Screenhouse water level decreases. As the water level in the safeguards bay decreases to less than 669'-
j 0", the supply from the Plant Screenhouse will then be augmented by the cooling water emergency intake line. At a water level of 669'-0" in the safeguards bay, the water level in the remaining areas of the screenhouse is at i
approximately 670'-0" to support adequate flow through the sluice canals. At 31,750 gpm, the time to drain the screenhouse to 670'-0" (not including the circulating water pipe) is 5.0 minutes. With the augmented flow from the cookng water emergency intake line (not including the circulating water pipe) it takes 10.7 minutes to drain the safeguards bay from 669'-0" to 662'-0". The circulating water pipes provide an additional 2.2 minutes of available water supply.
Therefore, the total time available using this methodology is 5.0 + 10.7 + 2.2 = 17.9 minutes 1
This assumes no reduction in cooling water system demand during the event.
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The effect of operator action to reduce the cooling water system flow demand is also evaluated. It is assumed that the cooling water system flow demand is reduced to 21,750 gpm at time 5.0 minutes. This represents the operator isolating the flow to Unit 1 and Unit 2 turbine buildings (the first two steps of flow reduction in the earthquake procedure). The time to drain the safeguards bay from 669'-0" to 662'-0" is increased from 10.7 minutes to 18.4 minutes. The total time available is increased from 17.9 minutes to 25.6 minutes. The additional flow reductions steps will continue to increase the total available time.
l.A.4. River elevation The U.S. Army Corps of Engineers (COE) was contacted in November of 1995 to determine the response of the Mississippi River to the loss of Lock & Dam 3.
They said that it would take approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for Pool 2 (upstream of Lock
& Dam 3) to equalize with Pool 3 (downstream). This results in a river elevation of approximately 669'-0". The river would remain at this elevation for about 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> (COE rough estimate) while the volume of water upstream flowed past the plant into Pool 3. Only then could river level decrease to the postulated low level of 666'-6" Conservatively, it is assumed that the river decreases to 669'-0" immediately following the design basis earthquake. From the testing of the emergency intake line, a river elevation of 669'-0" will provide 13,700 gpm through the line.
Therefore, it can reasonably and conservatively be concluded that operator action must reduce cooling water system flow demand to 13,700 gpm within the first three hours after an earthquake. The flow demand must ultimately be reduced to 11,600 gpm, which is the capacity of the emergency intake line at the assumed minimum river level of El 666' 6"
- 1. B.
Operator action To assure timely entry into the earthquake procedure, a dedicated operator will be assigned to monitor the seismic annunciator. This operator will be in addition to the normal administrative control room staffing. All licensed operators will be trained on this earthquake procedure prior to ascuming license duties after Unit 2 exceeds Mode 3 (Hot Shutdown).
A seismic event is detected by accelerometers. A deflection of 0.03 g starts the seismic recording system and activates the " SEISMIC EVENT" annunciator. The seismic annunciator is on the main control panel in the control room and has a unique horn.
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Placing a dedicated operator in the control room allows for immediate entry into the earthquake procedure. It will be performed simultaneously with the reactor trip procedure. For a design basis earthquake, there are no contradictions between these two procedures. As demonstrated in our validation tests, these two procedures compliment each other.
Also demonstrated in the validation tests, the earthquake procedure can routinely be performed within 9 minutes. During validation, all operators carried out the reactor trip procedure immediate actions prior to implementing the earthquake procedure. This provides reasonable assurance that operator action will be successful in reducing cooling water system flow demand within this shortened time frame.
Throughout this evaluation, river level is assumed to decrease immediately to 669'-O
Flow through the emergency intake line at this elevation is 13,700 gpm. Ultimately, river level is postulated to decrease to 666'-6", and emergency intake line flow to 11,600 gpm. Even though the first two major steps for isolating cooling water loads (both turbine buildings and all containment fan coil units) will reduce demand to less than 13,700 gpm, the operator will continue in the procedure to reduce demand to 11,600 gpm.
After the cooling water system flow demand is reduced to within the capacity of the emergency intake line, it is desirable to close the sluice gates. This action ensures the long term protection of the pumps from any debris that may be introduced into the plant screenhouse. This is a long term action and may be performed as personnel resources allow. This direction is provided to the operator in the Recovery Actions of the earthquake procedure. More discussion on the effects of debris is presented in a later section.
The effects of a single failure have been evaluated previously. Prairie Island is capable of maintaining both units in safe shutdown including the effects of a single failure.
However, the effects of a single failuro on reducing cooling water system flow demand within a reduced time frame are evaluated here. The reduced time frame is applicable to these interim actions. The most limiting single failure is one which prevents the control room operator from reducing the flow demand from the control room. This would be a safeguards bus lockout or a diesel generator failure. The result of the single failure is that multiple valves cannot be operated from the control room. The most limiting combination of valves is that which includes the motor valve which isolates flow to a turbine building. Because of the high flow rate through this valve, added to the other three valves without power, there are not enough other valves which can be closed from the control room to reduce cooling water system flow adequately. The contingency operator action is a local manual action to close the appropriate turbine building cooling water isolation valve. It is important to note that only this valve is Page 8
needed to be closed within the reduced time frame. Other control room actions in the procedure will be effective in reducing flow to less than 13,700 gpm.
The earthquake procedure has been revised to include the contingency action for locally manually closing either cooling water to turbine building isolation valve. The new revision has been validated using the simulator. The crew composition was the normal administrative minimum plus the dedicated operator. The sequence of outplant actions has been validated in the plant. This includes verification of accessibility to components, adequate emergency lighting, safety of personnel performing the evolution, and ergonomics.
The dedicated control room operator's function is to recognize the seismic annunciator and notify the Shift Supervisor. The Shift Supervisor directs the implementation of the earthquake procedure. The dedicated operator is then an additional resource to the control room crew. The procedure directs that the dedicated operator is to be sent to manually isolate the turbine building cooling water valve, if necessary. This eliminates the need for outplant operator notification. Because this operator is an addition to the control room crew, completing an outplant action does not reduce the control room j
crew to less than the administrative minimum.
An evaluation was performed to determine the time frame that would be available to complete this manual action. For this evaluation, it has been assumed that D1 emergency diesel generator is the single failure. Operator actions were assumed per the following time line. Times are approximate. At time 4 minutes, the motor valve for the Unit 1 turbine building isolation is attempted and is unsuccessful. The dedicated operator is dispatched to perform the local manual isolation. At time 5 minutes the Unit 2 turbine isolation valve is closed. At time 7 minutes, two containment fan coil unit outlet valves are closed. At time 9 minutes, two additional containment fan coil unit outlet valves are closed. At time 11 minutes, two additional containment fan coil unit outlet valves are closed. Because of the single failure, two containment fan coil unit isolation valves cannot be isolated from the control room. At time 13 minutes, one Unit 2 component cooling water heat exchanger inlet valve is closed. No credit has been taken for the Unit 1 turbine building isolation. The time calculated until the safeguards bay is drained is 27.1 minutes. Therefore, from the time that the turbine building motor valve is found not operable from the control room, until the action must be completed, is 23.1 minutes. The walkdown validation has determined that the isolation can reasonably be completed in 4.5 minutes.
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l I.C.
Procedure Verification and Validation Validation of the earthquake procedure, Rev.12, was performed during the last two weeks of February,1997 for the purpose of validating procedural guidance and demonstrating expected operator response times. This validation effort was in support of the subject license amendment request.
Subsequent to that validation, significant changes to the procedural guidance have 4
been made in an effort to enhance operator response times, and also to address the apparent need for operator action outside of the control room in the event of a limiting single failure scenario which renders one of the turbine building cooling water valves, two containment fan coil unit cooling water return valves and one component cooling heat exchanger inlet motor valve inoperable from the control room.
Specific procedural enhancements included assignment of a dedicated operator to provide early awareness of a seismic event, moving of a non-immediate step from the beginning of the procedure and placing it in a more appropriate position after the cooling water reduction steps, and removing one step entirely since it was redundant.
To address the limiting single failure scenario rendering four of the motor valves inoperable from the control room, procedural guidance was added to direct the dedicated operator to perform manual closure of the inoperable turbine building cooling water valve locally at the valve. With closure of this valve, reducing flow approximately 5,000 gpm, the inability to close the inoperable tu o containment fan coil unit valves and the component cooling heat exchanger inlet motrer valve does not interfere with the capability of the control room personnel to achieve the desired flow reduction using the remaining control room actions as specified in the earthquake procedure.
l.D.
Validation Results I. D.1 Control Room Actions A total of 13 validation runs were made using the simulator after the procedural enhancements and dedicated operator were incorporated. These runs resulted in an average time to reduce cooling water demand to less than 11,600 gpm of 7.1 minutes. This is a reduction from the average time of 8.1 minutes experienced during the previous runs. The time reduction is attributable to the addition of the dedicated operator to detect the seismic annunciator early in the scenario. (Average time to begin cooling water reduction steps in early scenarios Page 10
l was 3.9 minutes, compared to an average of 2.3 minutes during the 13 subsequent runs.)
Two significant facts support the conservatism of the validation effort. One is that the target flow rec 9 tion of the earthquake procedure is 11,600 gpm, yet the emergency intake pipe capability at the assumed river elevation is 13,700 gpm.
The other is that the simulator does not model the flow reduction of the Unit 2 containment fan coils, so an available flow reduction from the control room of approximately 4000 gpm is not modeled during simulator scenarios, even though the operators performing the validation addressed the steps.
1.D.2 Outplant Operator Actions l
A total of 21 walkdown exercises were performed to validate the dedicated operator action to locally close one of the turbine building cooling water supply valves. Ten were performed on MV-32031, and 11 on MV-32033. These exercises consisted of the following sequence of events.
- 1. Operator in control room was instructed to manually close the desired motor valve and stopwatch started.
- 2. Operator proceeded to the desired valve (choosing familiar route).
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- 3. Operator ascended scaffolding and got in position to engage declutching lever and rotate valve handwheel.
- 4. Operator rotated hand and arm in the motion of closing the valve 156 turns and stopwatch stopped.
The average time recorded for these 21 validations was 3 minutes and 37 seconds. The longest being 4 minutes and 30 seconds, the shortest being 2 minutes and 35 seconds.
These valves are stroked, using their motor operators, on a quarterly basis by surveillance testing and are thus assured to be free to operate smoothly.
To support this timed validation, an actual closure, by hand, of MV-32033 was performed to verify ease of operation of this style of butterfly valve with the actual manual operator gear ratio. The closure time was 80 seconds.
To support the ability of the operator to gain access to the motor valves for manual closure, scaffolding has been erected and evaluated for adequacy Page 11
4 following the design basis earthquake. Some modification to the scaffolding was made following the first two validations to improve the ergonomic conditions of the manual operation.
Emergency lighting has been installed to support the valve manipulation and scaffold access. The lighting is powered by seismically installed battery packs and directed appropriately to illuminate the scaffolding and the motor valves.
The lights will automatically turn on following loss of offsite power.
The valves and associated access scaffolding are located in switchgear rooms inside the seismically qualified switchgear corridor between the turbine buildings of the units. These rooms each have two entry doors from their respective turbine buildings and an interconnecting door. This provides three access points to each valve.
The operators participating in the validation were given the liberty of selecting their own pathway to the desired bus room. Two stairwells are available to provide basically equal length pathways to the switchgear rooms from the control room. These pathways have been walked down to verify they will remain relatively free from obstruction and adequately lighted following the design basis earthquake.
A comparison was made to determine if one pathway was preferable to another based on timed results. No significant time difference was identified so no preferred pathway is identified.
- 1. E.
Conclusions of the Cooling Water Load Management Evaluations 1
Conservative evaluations of the water sources available within the plant screenhouse i
show that at least 17 minutes are available for the operators to manage cooling water system loads. This time is significantly extended once the operator initiates the procedure. Validations of the procedure have demonstrated that the operators perform the procedure well within the available time. Therefore, the cooling water system loads can be managed in the time available using conservative assumptions on the available water supply.
i i
IL Potential Entrained Debris i
it is possible that debris may be introduced into the water that flows to the safeguards bay and is pumped through the cooling water pumps. The first source of debris that is Page 12
I considered is a mixture of rock, sand and gravel that would be introduced into the water in the plant screenhouse as a result of the seismic event. This would be a singular event. The second source of debris that is considered is a continuous flow of a water, sand and grass mixture into the plant screenhouse from the intake canal.
The first case is where debris is introduced into the plant screenhouse as a result of the seismic event. A postulated collapse of the intake canal banks into the intake canal could contain large rock, gravel, coarse sand and fine sand. The water volume into which this debris mixes is outside of the non-safeguards traveling screens. The screens do not have power to rotate nor water for backwash. The velocity of the water determines the type of debris that can be transported. The eight traveling screens are 11'-2" wide, and the water is 14'-6" deep (normal river level). If only 13' of water depth 2
is assumed, this is a flow area of 1144 ft. Using the maximum flow rate in the srceenhouse of 31,750 gpm, this results in a water velocity of 0.062 ft/sec. As the water volume is depleted, assuming the flow rate remains constant, the water velocity will increase. With only 25% of the original flow area available, the water velocity is 0.427 ft/sec. This very low water velocity can transport coarse and fine sand, but not gravel and rock.
The trays in the traveling screen have a 3/8" wire mesh. Each of the eight traveling screens is designed to pass 80,000 gpm (assuming clean and 13'-6" water depth). This is a water velocity of 2.56 ft/sec. After a seismic event, the very low approach velocity and the 3/8" mesh ensure that no debris exceeding the nominal design will be transported to the safeguards bay.
The second case is where there is a constant flow of water, laden with debris, into the plant screenhouse. If the plant screenhouse water level is maintained constant, then the full 31,750 gpm flow rate would come from the intake canal. To increase the water 2
velocity to 1 ft/sec, the flow area would have to be reduced to 70 ft. The nominal cross 2
sectional area of the canal is 1155 ft (110' across bottom x 10.5' normal level).
Therefore a 1 ft/sec water velocity is conservative. Again, gravel cannot be transported by the water. However sand and grass will be transported to the plant screenhouse.
Sand can pass through the traveling screens, but grass could mat and clog the screens. The screens can withstand 8' differential head by design. Greater than this value is assumed to cause the screen to fail and allow sand and grasses to pass through unfiltered. Once the water has passed through the traveling screens, the flow area will again increase and the water velocity will decrease. As the water continues to the safeguards bay, it passes through the sluice gates and sluice canal. The bottom of the sluice gates is 2' above the plant screenhouse floor. This physical barrier combined with very low water velocity will only allow sand and grass to enter the safeguards bay.
1 Page 13
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t As stated above in the Operator Action section, a step to close the sluice gates was added to the Recovery Actions of the earthquake procedure. This step is performed after the flow reduction has been completed. Closing the sluice gates provides I
additional assurance of a long term debris-free water supply to the cooling water pumps.
The evaluation of the second case is noteworthy. The continuous transport of water into the plant screenhouse provides additional time for the operator to reduce cooling water system flow demand. The flow provided does not negatively impact the operation of the safeguards cooling water system.
The safeguards cooling water pumps were procured as safety related, with seismic qualification. The specification required the pump to provide the design flow rate assuming the fluid is from the Mississippi River and strained through a 3/8" opening.
The above discussion demonstrates that the design assumptions are preserved. The cooling water pumps will pump water of the quality described above without damaging the pumps. To preclude plugging of cooling water system components (e.g., heat exchangers), there are strainers in the downstream piping. The strainers are safety related and seismically qualified. The differential pressure switch, backwash valve strainer motor are safety related and seismically qualified as are the power supplies.
Therefore, all downstream components are protected from blockage.
Thus, potentially entrained debris will not negatively impact the operation of the J
safeguards cooling water system.
1 Ill.
Low Probability of the Desian Basis Earthauake and other Assumed Failures Ill. A.
Initiating Event Assumptions An evaluation of the risk involved for interim plant operation under the proposed license conditions was performed. Many conservative assumptions have been made which eventually result in the loss of safeguards bay level. Due to the successive layering of these assumptions, the likelihood of the event is extremely low. The scenario is as follows:
A design basis seismic event occurs. For Prairie Island, this event has a magnitude of 0.12g. Typically, earthquakes of this magnitude do not cause major damage to man-made structures.
No credit for the survival of Lock & Dam #3 following the initiating earthquake is given. This is based on the lack of documentation of seismic qualification of the l
Page 14 i
l dam. Lack of documentation does not mean that the dam will fail during a seismic i
event with a probability of 1.0.
Complete, total loss of function of Lock & Dam #3 is assumed, such that the Mississippi River is assumed to fall to the assumed level,669' elevation, l
immediately following the earthquake. This provides the minimum flow conditions for l
the emergency intake line flow rate into the safeguards cooling water pump bay in the time immediately following an earthquake when operator actions are required to manage cooling water system loads. However, the U.S. Army Corps of Engineers j
has estimated that if Lock & Dam #3 were to fail catastrophically the river level will remain above elevation 669 for at least 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and it would take a total of approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> for the river to drop to its minimum level. Therefore, although river level would eventually decrease if Lock & Dam #3 is assumed to fail, there will be more flow available to the cooling water safeguards bay through the emergency intake line than is taken credit for in the operator action to manage cooling water loads.
No credit is given for the availability of the intake canal water volume. The STS evaluation is considered to be a best estimate evaluation. It is not the conservative i
calculation needed for a design basis evaluation. However, it provides reasonable j
assurance that intake canal bank failure causing complete blockage of the plant screenhouse will not occur with a probability of 1.0.
Even though the probabilities of these items are less than 1.0, since they have not been quantified, they have been assumed to be 1.0 for the rest of this discussion.
Ill.B. Probability of Design Basis Seismic Event Central to the probabilistic evaluation is the probability of occurrence of the design basis seismic event. NUREG-1488, " Revised Livermore Seismic Hazard Estimates for Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains," gives the annual probability of exceedance for selected peak ground accelerations for nuclear power plants east of the Rocky Mountains. The design basis seismic acceleration for Prairie Island is 0.12g. Using the NUREG-1488 data for Prairie Island, the annual mean probability of exceedance for the Prairie Island design basis seismic event is estimated at approximately 1.2E-4.
Note that this is only the probability that the design basis seismic event will occur; not o
the probability that the seismic event will occur, Lock & Dam #3 will fail, the Mississippi River will immediately fall to the assumed level of 669', offsite power, instrument air, and condensate storage tanks will be lost, and that intake canal walls will slough into the canal such that the entire plant screenhouse is blocked. As described above, there l
is reasonable assurance that the probability of a design basis seismic event with all of the assumed initiating conditions would be much lower.
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lil.C. Worst Case Single Failure Considerations l
Operator actions for implementing cooling water load management in the earthquake procedure can be performed completely from the control room. With the assumption of the most limiting single active failure (loss of either D1 or D6 emergency diesel generator), operator action from outside the control room would be necessary (to manually close the train-associated motor-operated cooling water isolation valve for turbine building loads). Using Prairie Island updated Probabilistic Risk Assessment models, the probability of the need for this outplant operator action was investigated.
The following key assumptions were used in the analysis:
Based on an estimated time frame of 9 months for resolution of the intake canal seismic analysis issue, the probability that the design basis seismic event will occur prior to resolution of the unresolved safety questions is 1.2E-4 (9/12) = 9E-5.
Complete loss of offsite power and loss of the instrument air system were assumed to occur immediately upon the seismic event. The unavailability of the condensate storage tanks does not impact the availability of D1 or D6 diesel generators.
No credit was given for any water from the intake canal.
The diesel generators (and their support systems) were assumed to be exposed to failure for one hour following the seismic event. Therefore, diesel generator failures were conservatively included in the analysis even if they occurred after the calculated latest time that the operator could isolate the valves for cooling water load management success.
No credit was given for any operator actions to restore power to the 4160V safeguards buses supplied by D1 and D6 (Bus 15 and 26, respectively), or for restoration or recovery of any failure involving the failed diesel generator or its support systems.
Maintenance activities (preventative and corrective) for the diesel generators and support system components were included in the unavailability modeling.
The results of this analysis indicate the conditional probability of failure or unavailabiiity of either D1 or D6 diesel generator following the design basis seismic event is approximately 1.1E-1 over the next 9 months. This is driven by the expected performance of the D1 and D6 diesel generator annual preventive maintenance procedures, and by random failures of the diesel generators to start and run.
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lli.D. Conclusion Therefore, the overall likelihood of the need for outplant operator action during the interim period with License Condition 1 is 1.2E-4 (9/12) (1,1E-1) = 9.9E-6. Again, this is based on the assumption of extremely conservative initiating conditions as described above. Accordingly, there is reasonable assurance that the actual probability is actually much lower.
l IV. Safety Evaluation Conclusions The NRC (Reference 1) concluded that taking credit for the non-seismic intake canal and operator actions following an earthquake involved unreviewed safety questions.
Analyses have been performed which conservatively determined that a minimum of 17 minutes are available for the plant operators to reduce cooling water system demand to match the emergency intake line capacity. NSP has implemented a procedure for 4
managing cooling water loads and has demonstrated it wili be implemented within the available time. License Condition 1, through the commitment to provide a dedicated operator in the control room, provides further assurance that the occurrence of an earthquake is recognized and accordingly the cooling water system loads will be i
reduced in the available time Therefore, the cooling water system loads can be managed on an interim basis in the time available using conservative assumptions on i
available water supply.
Evaluations have been presented in this submittal which demonstrate that debris which could potentially enter the plant screenhouse flow stream will not impair the ability of j
j the cooling water system to perform its safety function.
The probability (9.9E-6) of occurrence of the design basis earthquake with the worst case single failure and the other assumed failures is low. This provides additional assurance that operation with the proposed license conditions is safe.
NSP has demonstrated through the subject license amendment request and the supporting supplements, including this supplement that the requirements of all applicable regulations have been met. Therefore, the health and safety of the public is protected when Prairie Island is operated in accordance with this proposed license amendment with license conditions.
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DETERMINATION OF SIGNIFICANT HAZARDS CONSIDERATIONS The proposed changes to the Operating License have been evaluated to determine whether they constitute a significant hazards consideration as required by 10 CFR Part 50, Section 50.91 using the standards provided in Section 50.92. This analyses is p:ovided as follows:
inis supplement presents new conditions for plant operation, new evaluations as the casis for operations and new conclusions, therefore new Determination of Significant Hazards Considerations are included.
- 1. The proposed amendment will not involve a significant increase in the probability or consecuences of an accident previousiv evaluated Operation of the Prairie Island plant in accordance with the proposed changes does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Probability The accident of concern for this issue is a seismic event. None of the proposed changes can have any effect on the probability of a seismic event.
Consecuences Operator action is required to assure continued operation of the cooling water system following a design basis earthquake. Evaluations have been performed to determine the length of time available for these operator actions. The eartnquake procedure directs the operator actions to assure they are performed in an accurate, timely manner. This procedure was developed in accordance with human factors considerations and has been validated and verified with many plant operators to assure that it will be performed in the available time. All licensed control room operators will be trained on this procedure. A dedicated operator will be provided in the control room to further assure timely reduction of cooling water loads following an earthquake. These considerations collectively assure the cooling water system will perform its safety function following a design basis earthquake. Thus, use of operator actions within the time frame available does not involve a significant increase in the consequences of an accident previously evaluated.
Failure of the intake canal banks could create debris in the intake canal. The effects of this material on the operation of the cooling water system have been evaluated. The low flow velocities in the plant screenhouse intake bay would not Page 18
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transport materials large enough to harm the cooling water pumps. Furthermore, l
the plant screenhouse geometry would provide traps for materials prior to their l
reaching the cooling water pumps. The cooling water pumps and system are
[
designed to handle small suspended materials such as sand fines and screenings l
less than 3/8" diameter. Thus, the cooling water system will continue to perform its l
safety function following a design basis earthquake and a significant increase in the consequences of an accident previously evaluated is not involved.
in conclusion, the changes proposed by this license amendment request do not I
involve a significant increase in the probability or consequences of an accident previously evaluated.
2.
The proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed The cooling water system is provided in the plant to mitigate accidents and it is not a design basis accident initiator, thus the proposed reliance on operator action in the available time and consideration of bank failure effects do not create the possibility of a new or different kind of accident.
In total, this proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
The proposed amendment will not involve a significant reduction in the margin of safety The proposed license amendment does not involve a significant reduction in the margin of safety. The cooling water system loads would be reduced in accordance with a simple, straight-fonvard procedure. Procedure validations (timed testing) have demonstrated that a normal operating crew can implement the procedure within the available time. The licensed operators have been trained on the earthquake procedure and a dedicated control room operator will be present to identify occurrence of an earthquake and assist in the response to the earthquake.
Thus, the operator actions proposed in the license amendment do not involve a significant reduction in the margin of safety.
Postulated failure of intake canal banks have been considered for their potential to harm the cooling w Mr system. The evaluations show that the flow velocities are too low to cause trusport of materials which would harm the cooling water pumps.
The materials that could be transported are within the capability of the system to l
pass through without harm. Furthermore, the screenhouse geometry provides traps l
for materials which prevent them from entering the safeguards bay. Thus Page 19 i
postulated failure of the intake canal banks does not involve a significant reduction in the margin of safety.
Overall, this proposed license amendment with the proposed license conditions do not involve a significant reduction in the margin of safety.
Based on the evaluation described above, and pursuant to 10 CFR Part 50, Section 50.91, NSP Company has determined that operation of the Prairie Island Nuclear Generating Plant in accordance with the proposed license amendment request, as modified by this supplement with license conditions, does not involve any significant hazards considerations as defined by NRC regulations in 10 CFR Part 50, Section 50.92.
ENVIRONMENTAL ASSESSMENT NSP Company has evaluated the proposed changes and determined that:
1.
The changes do not involve a significant hazards consideration, or 2.
The changes do not involve a significant change in types or significant increase in the amounts of any effluents that may be released offsite, or 3.
The changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR Part 51, Section 51.22(c)(9). Therefore to 10 CFR Part 51, Section 51.22(b), an environmental assessment of the proposed changes is not required.
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