ML20052G121
| ML20052G121 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/10/1982 |
| From: | Blanchard D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20052G119 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8205140326 | |
| Download: ML20052G121 (41) | |
Text
{{#Wiki_filter:i, Ek UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) Docket No. 50-155-OLA CONSUMERS POWER COMPANY ) (Spent Fuel Pool ) Modification) (Big Rock Point Nuclear Power ) Plant) ) FURTHER TESTIMONY OF DAVID P. BLANCHARD ON CHRISTA-MARIA CONTENTION 8 AND O'NEILL CONTENTION IIIE-2 My name is David P. Blanchard. I am employed by Consumers Power Company as a Technical E"gineer at the Big i Rock Point Plant. I was responsible for writing one of the summary disposition affidavits concerning Christa-Maria Contention 8 and O'Neill Contention IIIE-2. My qualifications were presented as a part of that affidavit. In.this testimony I will answer the following Board questions with respect to this contention: l 1. How reliable is the remotely activated makeup water system which will be added to the spent fuel pool? How reliable does it need to be? How many gallons per minute will it be able to make up? 2. How reliable are the spent fuel pool l water level monitors which the applicant is planning to install? Is applicant required to install and maintain these monitors? $0'hpayK05000155 26 820510 7 \\ PDR 1
e ~i l 3. Are motor-operated valves MO-7064 and 7068 necessary to control containment pressurization? Are they qualified for high temperature, high humidity environment? 4. Will Zircalloy react with steam in a fuel pool which is boiling because its cooling system has failed? Will the reaction become self-sustaining? Introduction (Question 1) The first question is addressed to the design and operation of the remotely actuated makeup system installed to permit maintenance of fuel pool water inventory in the event containment is uninhabitable as a result of an accident similar to that which occurred at TMI-2. 1. How reliable is the remotely acti-vated makeup water system which will be added to the spent fuel pool? How reliable does it need to be? How many gallons per minute will it be able to make up? In ascertaining the reliability of the remotely actuated makeup system to the fuel pool, I will be referring to Figure I which is a simple line diagram detailing important features of the system. The figure shows the spherical containment and major equipment associated with the makeup system inside and outside containment. This figure is similar to Figure 2 attached to my summary disposition affidavit, except that it shows some additional details and corrects an error in the previous figure as to the location of valve MO-7066 (this error has no significance for the purposes of my l 1 l l I
g . previous affidavit). Like all line diagrams, Figure I omits many details of the systems portrayed. However, all important systems and components necessary for an understanding of this contention and my testimony are included. Summarizing the course of events during this accident described in my earlier statement, a loss of coolant accident occurring as it did at Three Mile Island results in the reactor coolant inventory being blown down out of the reactor into the containment building. As containment i pressure rises to 2.2 psig due to the LOCA, the enclosure spray valve MO-7064 automatically opens to condense the steam in containment and reduce containment temperature. After the water in the reactor vessel drops to a level less than 2'9" above the core, a signal is generated causing valves in the reactor depressurization system'to open, completely depressur-izing the reactor into the containment. As reactor pressure falls below 200 psig, motor-operated core spray valves MO-7051, MO-7061, MO-7070, and MO-7071 open automatically spraying fire protection system water on the core to cool it. Fifteen minutes following initiation of the incident, the operator opens motor-operated valve MO-7068 which provides additional containment spray to further condense steam and wash out any fission products present in the containment atmosphere. l The water entering containment through the opening in the primary system, the reactor depressurization system, I l
s _4_ the core sprays and ultimately the enclosure sprays eventually collects in the bottom of the containment building. There is a limit to the height to which water can be added to contain-ment in order to prevent excessive stress of the containment shell due to the static head of water. This limiting eleva-tion is approximately 23 feet above the bottom of containment. The operator is instructed to terminate the level rise in containment by placing a core spray pump in service and isolating hand-operated valves VFP-29 and VFP-30 approximately at the time the level reaches 14 feet above the containment bottom. This stops fire water addition to the containment and yet continues to keep the core sprays operating by recycling water from containment, cooling it with the core spray heat exchanger and pumping it back to the core. Motor-operated valve MO-7066 is opened to provide the cooling medium for the recycled water in the core spray heat exchanger. It is at this stage of the accident that water addition to the fuel pool begins. A portion of the water l drawn from the containment by the core spray pumps is routed l through valve VPI-18 to the fuel pool for the purposes of 1 l maintaining pool inventory. It is postulated that water loss l from the pool may be occurring as a result of pool boiling due l to the loss of the normal pool cooling loop during the LOCA. However, it should be recognized that the normal cooling equipment may in fact survive the LOCA. Further, for purposes of this analysis, we have ignored the effects of natural heat l
e s conduction through the pool walls and water added to the pool due to actuation of the containment sprays. It is possible that pool boiling would never occur. Discussion (Question 1) I will begin the discussion of the adequacy of the fuel pool makeup system by first addressing how reliable it is. There are only two active components in the core spray recirculation system either of which must actuate in order to add water to the pool. These are the two core spray pumps. As will be presented later in this discussion, the core spray pumps each have a capability great enough to provide suffi-cient flow to the core spray system to cool the core and to the fuel pool makeup line to cool the spent fuel. The two pumps are, therefore, redundant providing assurance that pool and core cooling can be maintained even if one of these pumps were to fail. These pumps are located outside the contain-ment. There are, therefore, no active components in this system which are required to operate in the accident environ-ment within containment. The power for each pump is supplied from a separate AC bus in the station power system. The normal power source for these pumps is off-site power. Either of these buses can be transferred to the emergency bus in the event off-site power is lost. The emergency bus receives its power from either of two redundant diesel generators, insuring the availability of power sources for operating these pumps.
a + It is concluded that due to this redundancy, failure of active components does not contribute significantly to the unavail-ability of the fuel pool makeup system. The remaining components in this system are all passive, that is they do not have to operate to place the system in service but merely provide a path for the core spray pumps to draw water from the containment and send it to the pool and core spray systems. These passive components include the suction and discharge of the pumps, the core spray heat exchanger, and the makeup line and valve to the fuel pool. Consideration has been given to precluding failure of these passive components in their design. I do not believe there are any mechanisms which will result in the failure of these components in providing water to the pool following a loss of coolant accident. For example, as can be seen in Figure I, the majority of the components in the fuel pool makeup system are located outside of the containment where there are no lines containing high energy primary coolant. Therefore, these components are not vulnerable to pipe whip or steam inpinge-ment or to the hostile environmental conditions inside containment following an accident similar to TMI-2.
- Further, the makeup line to the pool inside the containment is routed such that it does not pass in proximity to primary coolant system piping.
It is extremely unlikely that a failure of the primary coolant system leading to a LOCA, therefore, could l i
0 0 simultaneously cause a failure of fuel pool makeup system components. These components are located such that the drop of a cask or other heavy objects cannot simultaneously damage primary coolant system lines and components required for makeup to the fuel pool. Again because components in these two systems are not in proximity to one another, an event such as a cask drop does not pose a threat to being able to provide makeup to the fuel pool if required. It should be noted that there are hand-operated valves located throughout the core spray recirculation and fuel pool makeup lines. These do not need to operate to place these systems in service. However, the mispositioning of these valves could prevent flow to the systems within containment when placing the recirculation system in service. It is, therefore, important to consider the positioning of these valves in determining the reliability of this makeup i system. In most instances, these valves are only placed in positions which would prevent flow through the core spray pumps during refuelings or reactor shutdowns. The purpose of changing the position of these valves is to perform surveil-lance testing of core spray system components or to complete preventative maintenance on this equipment. l To insure these valves are returned to their correct i positions prior to returning the plant to operation, each surveillance procedure contains a step-by-step sequence of events which requires repositioning of these valves which is i
signed off by the operator as he corrects the valve line up. ~ In addition, tags are placed on each valve to indicate the valve has been placed in other than its normal position. The control room maintains a list of these tags and is not permitted to return the plant to operation until the valves are returned to their normal positions and these tagging orders are cleared. Additionally, system valve checklists (including one for the core spray system) are issued at the end of plant shutdowns such as refueling. Operators walk down the piping of the system and check the position of each valve and insures its position is as specified on his checklist. Two individuals complete each checklist providing a review on each other as the valve lineups are verified. Finally, valves important to safety such as those associated with the core spray recirculation system are locked in place such that they cannot be inadver-tently repositioned after checklists have been completed. There is then one final checklist which is performed to insure all valves requiring locks are in fact locked in the correct position. The extensive valve position control-s and lineup verifications performed provide significant assurance that mispositioned valves will not contribute significantly to the unavailability of the system. It is concluded that the likelihood of mispositioned valves due to maintenance and surveillance activities during refueling or shutdown does not contribute significantly to makeup system unreliability.
1 l One surveillance test is performed while the plant is at power which temporarily removes the core spray heat exchanger from service. During the period of time the heat exchanger is isolated for this test, the pumps of the core spray recirculation system will be unable to pump water through the heat exchanger. This' test is required by the plant Technical Specifications to be performed once each month subject to the limitation that the heat exchanger be out of service for a period of time no longer than four hours. Due to the short duration of the test (less than 1/2% of the time the reactor is at power), it is extremely unlikely a LOCA would occur while the heat exchanger is out of service such that the core spray recirculation system would be required. A hydrostatic test of the tubes in the heat exchanger is performed to insure their leak tightness should a LOCA occur and the core spray recirculation system be required. The core spray heat exchanger must be isolated to perform this hydrostatic test. This is accomplished by closing manual valve VPI-4. Like surveillance tests performed during refueling shutdowns, this test procedure contains a signed step-by-step sequence of events required to return the heat exchanger to service. On completion of the test, the operator returns and locks VPI-4 to its normal position using this procedure again assuring the heat exchanger is available for operation. n.-., w
' The performance of this surveillance test does not mean that the core spray recirculation system will be ~ unavailable should it be required to recirculate water to the core spray system or makeup to the pool. Recirculation is not required until the containment has filled sufficiently, and its is estimated that the earliest this can occur is four to five hours following a major LOCA. The valve realignments required to return the system to normal are all external to l l containment and entry into the containment environment is not required to return the heat exchanger to service. Moreover, the operator is instructed to terminate l the test and return the heat exchanger to service by opening l VPI-4 any time a reactor trip occurs during the performance of the test (regardless if the reactor trip is due to a simple transient or major loss of coolant accident). This insures the system is returned to service prior to any indication that I it will be needed and before radiation levels become prohibitive should severe core damage occur. It is concluded that valve mispositioning and existing surveillance testing of components in the core spray recirculation system are not significant contributors to the unavailability of the makeup system to the fuel pool due to the limited time over which these tests are performed and the ability to return the system l tc service prior to its being required. The reliability of the makeup system ultimately rests with the availability of sufficient water in the
O
- containment.
As stated before, a minimum level of 14 feet of water above the bottom of containment is required to achieve core cooling by the core spray recirculation system. The volume of water necessary to raise the elevation to this level has been determined to be approximately 260,000 gallons and it is necessary, as explained below, to attain this elevation within 1400 hours following the loss of the pool cooling system. It is therefore required that we insure a flow rate into containment of at least 3.1 gpm following a LOCA. The fire protection pumps which supply core spray and enclosure spray water are overwhelmingly sufficient in supplying the required flow to containment in that they are each capable of more than 1000 gpm. Additional diversity has been designed into this makeup system in that the operator has the option of adding water to the pool through this makeup line even if neither of the two core spray pumps is running, even if the core spray I heat exchanger is isolated and has been realigned for service and even if there is not water in the containment. The operator may accomplish this activity by opening motor-1 operated valve MO-7072. MO-7072 is a DC powered valve inde-l pendent of off-site or on-site AC power sources. It has a control switch located in the control room or it can be manually operated from outside the containment. Opening this valve permits fire protection water to be routed to the fuel pool from the fire pumps. This alternate method of water l
. addition is equally reliable as using the core spray recirculation system. As active components, the fire pumps are redundant to each other. One pump is electric and is permanently power from the emergency bus. The other pump is diesel driven and is therefore independent of off-site or on-site power sources. The passive components associated with the pump discharge piping, the yard piping and the piping up to MO-7072, are all external to the containment and cannot be damaged by pipe whip due or steam inpingment to high energy lines or heavy object drops which precipitate a need for the fire system. Valve alignments are subject to the same tagging and procedural controls as described for the post incident system. It is concluded that the fire protection system through the use of MO-7072 enhances the reliability of the makeup system to the fuel pool. Having established a basis for assuming the makeup system is in fact reliable, I will now attempt to establish requirements determining the degree to which the system should be reliable. I will then demonstrate that this system exceeds those requirements. The most commonly reference regulation with respect to system reliability is 10 C.F.R. Part 50, Appendix A. Although it was promulgated after Big Rock Point Plant was built and licensed, Appendix A establishes general design criteria with which engineered safety systems at Big Rock
. Point Plant may be compared. An important concept expressed ~ in Appendix A is the single failure criterion. Paraphrasing Appendix A, a single failure is an occurrence which results in the loss of capability of a component to perform its intended safety function. Fluid systems, such as the one designed to make up water to the Big Rock Point spent fuel pool, are considered to be designed against single failure if a single failure of an active component does not result in the loss of the system to perform i,ts intended safety function and if passive failures have been considered in the design of the system. (Criteria for design against passive failures in fluid systems, according to Appendix A, is under development.) The following discussion analyzes the makeup system against this standard. Recall from the discussion of the reliability of the core spray recirculation system that the only two active components in the portion of the system leading to the fuel pool were the core spray pumps. As either of these pumps is capable of providing core spray flow, as they are redundant to each other, and as they can be powered from the emergency bus during a simultaneous loss of off-site power with the LOCA, the core spray recirculation and spent fuel pool makeup system is not subject to single active failure. Recall also, that a detailed discussion of the potential for passive failure of the makeup system was presented and that the line is designed and routed such that mechanistic failure of passive makeup system components
. occurring at the same time as the LOCA could not be identi-fled. Finally, recall that a reliable alternate means of providing spent fuel pool cooling is available through the use of fire pumps and motor-operated valve MO-7072. It is con-cluded that the fuel pool makeup system not only complies with but also exceeds the single failure criterion as defined in 10 C.F.R 50, Appendix A and therefore clearly has adequate reliability. The remaining outstanding question in insuring the l reliability of the makeup system is whether or not it is l capable of delivering sufficient flow to prevent uncovering the fuel in the spent fuel nool due to boil-off. In answering the Board's question with respect to the rate at which water can be made up to the pool, I will first establish what rate i is required. l In my previous affidavit, I presented an analysis l which indicated that the maximum rate at which water could 1 boil from the pool following c loss of coolant accident was 2 gpm. This agreed quite well with the Staff's independent analysis; the results which were presented in their Safety Evaluation Report (SER) of May 15 1981. Unfortunately, I made a mistake in using the two gpm rate in my previous affidavit. This two gpm rate derived by both the Staff and me is applicable for the Big Rock Point spent fuel pool filled to capacity with the most recently discharged fuel bundles having l
. been removed from the reactor for several days, not one month as I said in my affidavit. The correct maximum decay heat generation rate one 5 month after refueling discharge is approximately 6 X 10 Btu /hr for a boil-off rate near one gpm; and the correct amount of time to uncover the spent fuel is nearly 1400 hours rather than 700 hours as originally stated in my affidavit. I should have used 1400 hours in my affidavit. This error is in the conservative direction and does not affect the conclusions in my affidavit. I should also note that based on the correction I have made to the heat generation rate, the time for pool boiling shown on page 3.4 of the Staff's SER should read 144 hours rather than 72 hours. An additional analysis was presented in our applica-tion dated April 23, 1979 indicating that a maximum boil-off rate of 11 gpm would occur if the fuel pool were filled to capacity with the full core discharged to the pool two days following the last reactor shutdown. (This rate was amended in a submittal dated October 16, 1979 to a value just under 8 i gpm.) Again the Staff presented results of their independent analysis indicating a maximum expected boil-off rate of 9 gpm, again in reasonably good agreement with Consumers Power Company's values. It will be assumed for the rest of this discussion that during plant operation it is necessary to have a makeup capability of 1 gpm and during refueling operations 9 gpm. t
, In the Board's Memorandum and Order (concerning Motions for Summary Disposition) the Board stated "We fail to understand why applicant promises only 'more than two gallons per minute' in its new makeup system". As the discussion below indicates, the makeup system will in fact supply a minimum of 13 gpm. My statement that the makeup system will supply "more than two gpm" therefore could have been stronger. However, the contention which I was addressing, refers to "an. accident similar to TMI-2". For such an accident (which cannot occur while the reactor is shutdown and its core is in the spent fuel pool), the maximum makeup flow needed in the spent fuel pool is only one gpm (at the time I wrote the affidavit, I thought 2 gpm was the correct value). Therefore I confined my statement to what I thought was necessary to answer the contention. In designing the makeup line to the pool, Consumers Power Company G:c'_ standard hydraulic techniques to determine the flow rate through the line under a variety of conditions using the following algorithm: 2 l bg = ( f - + k ) b PV D 2 e gc Where /g = Differential pressure across a given section of pipe 'L = Length of flow path
. D = Equivalent diameter of flow path e f = Friction factor (dependent on whether is lamanar or turbulent and fluid velocity) k = loss coefificient p = Fluid density V = Fluid velocity g = Conversion factor This algorithm is incorporated in a computer program entitled FLOWNET which was developed by MPR Associates, Incorporated for Consumers Power Company. FLOWNET has been used by Consumers Power Company to establish the adequacy of core spray and enclosure spray flows following postulated loss of coolant accidents at the Big Rock Point Plant. It was used to design the fuel pool makeup line by insuring adequate flow to the pool could be obtained without diverting water such that the adequacy of core spray flow to the core during recircu-lation was jeopardized. j Two models were developed which included the makeup line and run with FLOWNET: the core spray system as it would be lined up for recirculation following a LOCA and the fire protection system as it would be lined up to feed the makeup line through motor-operated valve MO-7072. Several cases were l run assuming various configurations of the core spray and fire L protection systems to insure adequate flow to both the fuel l pool and each of the two core spray lines. Cases were run l l l l l l
. which included single failures of components in the core spray system in accordance with 10 C.F.R. 50, Appendices A and K. To verify that the analytical techniques of FLOWNET were accurate and that the pool makeup line had been modeled accurately, a series of flow tests on the system were performed. It is not possible to test the core spray recircu-lation system as it would actually work following a LOCA (i.e., the containment cannot be filled with water, fire water injected to the reactor, or enclosure sprays wash down equip-ment in containment). However, it is possible to verify the cod,e using the valve configuration as it would occur if fire water were to be added to the pool through MO-7072. Several tests were performed verifying the results. Tests were varied to measure the flow to the pool with and without flow to the core spray hea't exchanger. The results of the FLOWNET analysis are as follows: Recirculation Cases (No Single ?ailures) 1. Nozzle Flow 386 gpm (296 gpm required) Containment Spray Flow (MO-7064) 79 gpm ( 50 gpm required) Fuel Pool Flow 14 gpm ( l gpm required) 2. Sparger Flow 342 gpm (292 gpm required) Containment Spray Flow (MO-7064) 96 gpm ( 50 gpm required) Fuel Pool Flow 17 gpm ( l gpm required)
- Recirculation (With Worst Single Failure - MO-7068 Open) 1. Nozzle Flow 348 gpm (296 gpm required) Containment Spray Flow (MO-7064) 70 gpm ( 50 gpm required) Containment Spray Flow 4 (MO-7068) 61 gpm (No requirement) Fuel Pool Flow 13 gpm ( 1 gpm required) 2. Sparger Flow 303 gpm (292 gpm required) Containment Spray Flow (MO-70 64 ) 67 gpm (No requirement) Containment Spray Flow (MO-7068) 100 gpm (No requirement) Fuel Pool Flow 14 gpm ( 1 gpm required) ~ Fire Protection System Case (No LOCA; one fire pump) 1. Fuel Pool Flow 24 gpm* ( 9 gpm required) 2. Fuel Pool Flow 22 gpm* ( 9 gpm required) Core Spray Heat Exchanger Flow 400 gpm* (No requirement)
- Verified by test.
t
. Reviewing the results, in no case during recirculation are the flows to the core spray lines (sparger or nozzle), the containment sprays or fuel pool makeup line less than that required to perform their functions, including cases in which the worst single failure has occurred. Conclusions The makeup system to the fuel pool is reliable. It is not subject to single active failures. It has been de-signed to avoid mechanisms which ceuld cause failures of passive components within the system at the time the system is needed. It complies with the general design requirments of 10 C.F.R. 50, Appendix A. Water can be added to the pool from either the core spray pumps or the fire protection pumps reflecting a defense in depth approach in its design. Hydrau-lic analyses indicate that flow to the pool is approximately thirteen times greater than required to makeup for boil-off assuming a TMI-2 event were to occur at Big Rock. The l analyses further indicate addition of the line to the core l spray system does not jeopardize required core spray or enclosure spray flows during recirculation. The spent fuel pool makeup line is adequate to protect the integrity of the spent fuel following an accident which prohibits entry to the containment. I 1
- Introduction (Question 2 - Pool Level Monitor) The second Board question conerning Christa-Maria 8 and O'Neill IIIE-2 pertains to the monitoring of spent fuel pool level. 2. How reliable are the spent fuel pool water level monitors which the applicant is planning to install? Is applicant required to install and maintain these monitors? By letters dated January 16, 1980 and May 15, 1980, Consumers Power Company committed to the installation a spent fuel pool level monitor, qualified for the LOCA environment, with readout in the control room. This commitment is reflected in the Staff's request for summary disposition. The commitment was made.in response to interrogatories concerning the ability to monitor pool level in the event of an accident which prohibits entry to the containment. The commitment to install pool level instrumentation was made prior to designing the remotely actuated fuel pool makeup line or determining the manner in which it would operate. Installation of the remote makeup system has now been completed and the manner in which the new pool level instrument interfaces with the makeup system can be examined. Installation of the pool level instrument is also nearly complete and will be operable prior to beginning installation of the new fuel racks'. Again, following a TMI-2 type event at Big Rock Point the containment will fill with water up to the point
. that core spray recirculation can take place before the fuel in the pool becomes uncovered duc to boiling. The operator will start the core spray pumps recyling water from the bottom of containment back to the reactor and the fuel pool. Water flow to the pool will take place at a rate more than suffi-cient to overcome water loss due to boiling and as a result the pool will begin to refill. The reliability of this makeup system was estab-lished in response to the Board's first question. Based on this reliability, the ability of the makeup system to more than overcome pool losses due to boiling, and the fact that this makeup will occur essentially automatically following the initiation of core spray recirculation, the benefits of the new pool level instrumentation following a reactor accident which prohibits access to the containment appear to be minimal. Nevertheless, Consumers Power Company did commit to install this instrumentation as a part of this proceeding. The instrumentation necessary to provide this information to the control room was on hand at the time the commitment was made. Consumers Power Company will, therefore, follow through on its commitment. The instrument is a Rosemount differential pressure transmitter, model 1152. It is mounted over the northeast corner of the pool and has an air-filled tube mounted below I the surface of the pool. Changes in water level result in a corresponding change in the pressure of the air in the tube. I
< This change in pressure is transmitted to an indicator on the, main console in the control room which is graded in feet of pool water. The power supply is from an instrument panel normally supplied by off-site power throught he station power system. In the event that off-site power is lost, the power source for this panel is automatically transferred to the emergency bus supplied by the diesel generator. Pool level indication is available to the operator under all conditions in which eithcc off-site or on-site AC power is available. The instrument is qualified for LOCA conditions as required by IEEE-323-1971. The radiation dose rates for which the transmitter is qualified are those determined for a 100% core damage situation in accordance with NUREG-0737 and Regulatory Guide 1.3. The instrument is qualified as seismic Class I in accordance with IEEE-344-1975. Conclusions (Question 2 - Pool Level Monitor) The new pool level monitor is required by Consumers Power Company Commitments date January 16, 1980 and May 15, 1981. The benefits of the instrument following a reactor accident which prohibits access to the containment appear to be minimal. The instrumentation does, however, provide direct information to the control room operator on the status of the spent fuel pool which was not readily available to him prior to this time. The instrumentation provides backup indication to existing methods of assuring adequate paol cooling. The instrumentation is reliable in that it is qualified for the _y-9 m
. LOCA environment, is seismically qualified and is powered from reliable off-site and on-site power systems. Installation of the pool level monitor provides additional operating informa-tion in the control room in ensuring satisfactory operation of the spent fuel pool system. Introduction (Ouestion 3 - Enclosure Spray Valvec) The third issue admitted by the Board is related to the use and qualification of motor-operated valves that control flow to the enclosure sprays following a loss of coolant accident: 3. Are motor-operated valves MO-7064 and 7068 necessary to control containment pressuri-zation? Are they qualified for high temperature and high humidity? Discussion (Question 3 - Enclosure Spray Valves) Valves MO-7064 and 7068 control the containment spray at the Big Rock Point Plant. The containment sprays serve two functions, namely, they reduce containment pressure and washdown iodine that may be present in the containment atmosphere. Neither the containment sprays nor motor-operated valves MO-7064 and 7068 is necessary to control containment pressurization. Following a primary coolant line break, the contain-ment pressure will rise to 2.2 psig at which time motor-operated valve MO-7064 will open automatically providing flow to a set of containment spray nozzles. This occurs very early in the accident. If the break is in a steam line (as opposed
- to a line carrying water) the containment sprays are necessary in order to keep the containment temperature below 235 F. This is the temperature on which the environmental qualifica-tion of Big Rock Point electrical equipment required following a loss of coolant accident is based. Approximately 15 minutes later, the operator is required to open MO-7068 providing sufficient flow for iodine washdown. (This washdown function is not essential for safe recovery from such.an accident.) MO-7068 can also be used as a redundant backup to MO-7064 in providing spray flow to reduce containment temperature. The containment is designed to withstand a pressure up to 27 psig. The amount of energy in the primary system coolant is such that there is no postulated loss of coolant accident which can result in a containment pressure this high. I have attached a plot of containment pressure versus time from a Consumers Power Company submittal dated March 15, 1982, which demonstrates the adequacy of the containment design pressure with respect to several postulated primary coolant line breaks which release primary coolant and steam to the containment (Attachment 1). Once it begins, the fuel pool is another source of steam which can result in a containment pressure rise. Recall that following loss of the normal source of fuel pool cooling that the pool will not begin to boil for approximately 140 hours. Note from the attached graph of containment pressure versus time that the pressure due to steam condensation on cool containment surfaces and enclosure spray operation will fall to near ambient conditions
. long before the time at which pool boiling begins. The contribution to the containment pressure rise of steam from the fuel pool, therefore, does not add to that which results from the reactor LOCA. Motor-operated valves MO-7064 and MO-7068 are qualified on an interim basis
- for the high temperature, high humidity environment created by a LOCA.
I have attached documentation also from the March 15, 1981 submittal elaborating the environmental conditions for which these valves are qualified (Attachment 2). This documentation is required by the Big Rock' Point Technical Specifications and NUREG-0588 to insure the reliability of this equipment following a loss of coolant accident. This documentation specifies the worst accident environment motor-operated valves are expected to see and provides information as to the quali-fication of this equipment in this environment. The NRC and their consultants did not at one time consider these valves to be fully qualified for this environ-ment. The documentation available for this equipment did not
- According to the Big Rock Point Plant Technical Specifications, the interim period for which continued operation is justified, based on the environmental qualification information described in this testmony, expires June 30, 1982.
This date is based on a deadline established by the Commission in its decision in the matter of Petition for Emergency and Remedial Action CLI-80-21, 11 NRC 707, 714-715 (1980). The Commission is presently considering a proposed rule which would extend this deadline to the second refueling outage following March 31, 1982, with further extensions possible based on a showing of good cause.
27 - meet all NRC Staff guidelines with respect to radiation and aging. Therefore, the NRC Staff required justification before it would allow Consumers Power Company to continue plant operation. Consumers Power Company provided this justifi-cation in Attachment 2 by demonstrating that the equipment is actuated early during an accident before the environment has degraded significantly and in fact is located in one of the environmentally milder areas of containment. The NRC accepted our interim justification by letter dated March 19, 1982 which is presented as Attachment 3. It can be concluded that suffi~- cient justification exists to believe these valves will function properly when called upon to do so. Following the LOCA at the time the pool begins to boil, the containment temperature and pressure will have returned to near normal and stem from the pool will being entering the containment atmosphere. As the environment created by pool boiling occurs very gradually (at a boil-off l rate no greater than 1 gpm), the conditions in containment 1 will be relatively mild compared to that created by the LOCA. Thee is little reason to assume containment sprays could not be used to condense the steam *rcated by the pool boiling. Therefore, it was not unreA On 51e for the NRC Staff to take credit for the operatior, at ...eAe sprays in its summary l disposition motion. But as it turns out, containment spray actuation will not be required at this stage of the accident. Recall that the makeup rate to the pool is greater than the 1 gpm required by boil off. Also, as time goes on
. the rate at which this makeup is required will decrease due to the gradual reduction in the rate that decay heat is being generated. As makeup water is added to the pool, the pool will stop boiling (if it is already boiling where the ma :w? waste system is turned on) because the cold water being added will more than compensate for the heat being generated by the fuel. As the pool level rises it will over flow the surge tank and flow back to the bottom of containment, to be drawn from the containment and recooled in the core spray heat exchanger before returning to the reactor or th9 pool. The pool will not resume boiling once this cooling path is established. Based on simplistic, but very conservative energy balance calculations, it can be shown that once core spray recycle has started, the time over which steam is generated by the spent fuel pool boiling is limited to 33 hours. The total amount of steam generated at a rate of one gpm over this period is no 6 more than 1.7 X 10 lb which is less than 20% of the mass of the primary coolant inventory which gets discharged to the containment during LOCA events. No credit is taken in this analysis for condensation of the steam as it is gradually j released from the pool. Therefore, it is obvious that the steam generated bl a boiling spent fuel pool will not overpressurize the containment under the circumstances I described above. 1 i
4
- Conclusions (Question 3 - Enclosure Sprays)
Enclosure spray valves MO-7064 and 7068 are not required to limit containment pressure resulting from steam entering the containment due to a loss of coolant accident or pool boiling. These valves are qualified for a high temperature high humidity environment on an interim basis and are used primarily to limit containment temperatures during a LOCA. There is no postulated LOCA which can cause the containment pressure to rise to its design value of 27 psig, even if containment spray valves MO-7064 and 7068 were to fail. The makeup system to the pool is designed such that heat generated by fuel in the spent fuel pool is ultimately removed by the core spray heat exchanger thus terminating boiling in the pool. Therefore steam from the spent fuel pool will not overpressurize.the containment. Introduction (Question 4 - Zircalloy Steam Reaction) The fourth issue admitted by the Board dealt with the possibility of a Zircalloy steam reaction in the spent fuel pool and its contribution to the rate at which boiling i occurs. 4. Will Zircalloy react with steam in a fuel pool which is boiling because its cooling system has failed? Will the reaction become self-sustaining? Big Rock Point fuel cladding is made of Zircalloy l l which can react with steam at high temperatures. The Zircalloy steam reaction is exothermic and the rate at which ( i I
- it occurs is highly dependent on the temperature at which the reaction is occurring.
The temperature dependence is exponen-tial in nature and the reaction ~ rate becomes significant at temperatures on the order of 2200 F. Elevated cladding temperatures such as this cannot be attained in the Big Rock Point spent fuel pool unless the fuel is allowed to become uncovered. The makeup system described in response to Question 1 (makeup system relia-bility) was shown to be sufficiently reliable that the boiling could not result in uncovering the fuel bundles. The maximum temperatures at which Zircalloy will be-in contact with water or steam will, therefore, be limited to the temperature at which boiling is occurring. The maximum possible coolant temperatures occurring in the Big Rock spent fuel pool have been-determined to be no greater than 237 F (Testimony of Daniel A. Prelewicz, dated May 7, 1982). This temperature is only a fraction of that-necessary to result in a significant Zircalloy steam reaction or to have it become self-sustaining. Indeed, boiling is occurring at all times in l the reactor during power operation in the Big Rock Point BWR. A coolant temperature on the order of 580 F exists during power operation with no observable self-sustaining Zircalloy steam reaction occurring. 1 i l l
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m 4 ATTACIIfirNT 2 + 'l EOtNPMENT OUALIFICATION REPORT l Owner: Consumeta Power Company Component Sleet Ho: I )! Facety: R10 Rock POIlff W: Docket: 50-155 Dele: ENVIn0NMENT DOCUMENTAit0N nEFERENCES OURrICADON OUTSTANDedG EOUW' MENT DEpHON p,, ,w ggggggnt Ousenceman Accident M METHOD HEMS i; system. Opweano Section Test and l {; Post-Incident Ih* 30 Days Sheet 2 II.D Sheet 2 Evaluation ti 'l F**l l D. Hu"**r Tempe'e8** i; MO-7061s (*FI Section Test and 235 Sheet 2 II.E 1 Evaluation Component, jj Valve Actuator eveesw. Section Test and Manussemaw: N 1:1.7 Sheet 2 II.E 1 Evaluation 1 Rotork f Modse ma=6er N Section Test and !) 11: A Syncroset Hua*mylst 100 Sheet 2 II.E 1 Evaluation - il e! Pwctmse Order Number S/N 53601: Section Test and l Lake Water Sheet 2 II.E 1 Evaluation i: runctioniservice: Containment nessman Section Test and y Spray Valve (ned 5 7 3 x 10 Sheet 2 II.E iheet 2 Evaluation Aceweer. Seac: Ashe Section 1heet 2 & Test and 1:0 Years + TDCA Sheet 2 II.E lection IV Evaluation q toconon-Q Containment subrawsonc* Not Subject Elevaton~ 625 To Submergence ~ 1 Hood Level 596 tieveson Ansove riood tevet: Yes: X No: i DOCUMENIATION REFEDENCES notes I Report F-C :121s performance qualification tests of 1. four valve motor operators. OS -8 I Hotork Test Report No TR :22, llent Aging Class B, NA2 2. L motor for "Outside Containment Safety-Related Duty." n 0 0W58 0t A
I EOUIPMENT QUALIFICATION REPORT Owner: Consumers Power Company Component Simot No.: ll FacEty: Birt Rock Ibint Revision. Docket: 50-155 Date: l !j DOCUMENTATION REFERENCES (cont) NOTES (cont)
- i 3.
Letter Hotork Hila to CP Co JLK dated October 17, it 1980 with attached summary for "A" rance actuators, ii la. Rotork Test No N/lis/2 " Actuator Radiation Subjectlon." 1 5 Limitorque Qualification Test Reports No B0003
- i j and No B0009 i
- I e
- 1 a
u li lI 'i G-0658-024
l 89 COMPOND4T ( valve Actuator Mfr Rotork Model: 14A Syncroset Plant I D NO MO-7064 FUNCTION This de valve actuator operates the primary Containment Spray line isolation valve which depending on its position, permits or prevent the flow of water from the line to the primary Containment Spray Nozzle. Actuation occurs on high containment pressure of 2.2 psig. DISCUSSION I This Rotork actuator contains a de motor with Class B insulation and . weatherproof enclosure..The actuator was installed in the Plant in 1970. Report F-C4124 " Performance Qualification Tests of Four Valve Motor Operators" subjected an exact type of actuator and motor to.a 36-hour LOCA simulation. The simulation followed the LOCA envelope curve for this period of time. The' test also incluced 24 continunus hours oi spraying for the first 24 hours'and then intermittent 1 hour spraying at hours J7, 31 and 35. Relative humidity was maintainec at or near 1004 by (1) use of saturated steam to obtain initial 0 temperature and pressure rise to 240 F, (2) use of fine mist water over the test specimen, -(3).use of saturated steam ejections to maintain chamber temperatures. Radiation and thermal aging were not included as part of the test procedure. Rotock has conducted testing for aging qualification. Rotork Technical Report TR422 subjected an NA type actuator with a Class B insulated 0 motor to an aging test which simulated 48 years at 60 C. The unit in the 0 0 test was placed in a furnace for 100 hours at 180 C. 48 years at 60 C was 0 determined by the 10 C rule. TR422 did not include a LOCA simulation, and although an NA type actuator was used in the test the only difference between the NA and 14A units, according to Rotork, is the motor hoJsing material, Cast iron versus cast aluminum. The results of the test, Rotork states, and the electrical characteristics will be exactly the same for the 14A as the NA type units. Rotork Test No N/14/2 used a prototype unit built for material evaluation for the Standard "A" and NA1 components. The actuator to MO-7064-being a Standard "A" type unit. The test results showed the "A" range components capable of withstanding 30 megarada during their 40-year life, according to Rotork. The test unit, however, was equipped with a Class H insulated motor. Limitorque actuators (Report B0003) with Class B, ac motors have been i irraciated to 20 megarads and been subjected to outside containment HELB - conoitions for a 16-day period anc have operated satisfactorily. Another Limitorque test (Report B0009) subjected a Class H, de motor to 10 megarads of gamma radiation and followec with a 25-hour HELB, simulation. The unit was 0 also preaged at 180 C for 100 hours ano satisfactorily passed the test. Rev 3 Doc ID 2240A-211A 3/15/82 (t e y
90 l The 30-day radiation dose given on the qualification sheet can be divided in / half as the actuator is mounted adjacent to a 3.5-foot thick, concrete wall. 5 An integrated dose of 3.7 x 10 rads is within the qualification dose of most materials. A one-day dose of 4.9 x 105 halved giving 2.5 x 105 is what the actuator will see during the time required to operate. This is also well within almost any material radiation resistance. l CONCLUSICN Although radiat on and thermal aging has been performed on similar type actuators, and therefore the probability is very high that the actuator in question will survive a DBA in an operable condition at the end of its qualified life, this component does not currently satisfy all the DOR guideline requirements. In order to fully comply with these requirements, a i more detailed analysis of the radiation withstand capability will be l performed, an aging analysis will be c'ompleted, and a qualified lifetime will be establishec. JUSTIFICATICH FOR INTERIM OPERATION Interim operation of this component is justified in lieu of. qualified equipment based on the following FRC Appendix A criteria 3 The unqualified equipment wil'1 have performed its safety function prior I to failure. As outlined in the surmary section of the discussion on Containment Atmospheric conditions, for steam line breaks requiring containment ,s,. spray, actuation occurs shortly following the event, when containment-environment has not degraded significantly (Reference Attachment II). MO-7064 is not located in the steam drum cavity, the area which experiences the temperature rise during the steam line break. Since the valve operator has limited exposure to the harsh environment during this event, failure is not expected. Rev 3 Doc ID 2240A-211A 3/15/82 i 9 e - -~ ,e- .~~.e._ ~
90 Sheet 2 (Contd) In summary, based on the testing of similar type units for radiation and thermal aging and the test conducted by CP Co, the actuator is expected to operate for the required one-day period. After one day, the valve will not be required to operate and will remain closed. The motor starter is installed in the station power room outside containment and will not be subjected to.a harsh environment; therefore, a misoperation due to failure of the starter is not a credible failure. The actuator is considered acceptable for use. i RESPONSE TO TER QUESTIONS A more detalied analysis of the radiation withstand capability will be performed. In addition, an aging analysis will be completed and a qualified lifetime will'be established. I s l i l 1 rp1080-0513b-63 Rev. 2 9/3/81 (
l I L EOUIPMENT DUALIFICATION REPOftT Component Sheet No: Focaly; n1G IKM:K TVIlfr fievision-* Docket: 50-155 t ENYWIDNMENT DOCUMENTATION nEFEnENCES g3, E""I E#" per.,,,ei., Accident ov'encamori Accident oud MEntoO HEM system: Op*'**'s Ibst-Incident TW Section Test and 30 Days Sheet 2 II.D Sheet 9 Evninntinn PlantI D. W 1ernpereIure Section Test and WT0@ l*N 235 235 II.E 1 Evaluation ca_. :, Valve Actuator
- presour, y,ssA, Section Test and 1:17 1:1.7 II.E 1
Evaluation Lismitorque Modd Haaber Section Test and' 34A-00 Hu"*8'r I'l 100 100 II.E 1 Evaluntion Purchase Order Neuber-Section Test and Lake Water Tap Water II.E L Evaluntion FunceorifService: Containment ninnemaa Section Test and Spray Valve 1 "
- 81 5
7.3 x 10 Sheet 2 II.E
- 1heet 2 Evaluation on,,,,
Section 1heet 2 & Test and Is0 Years + IOCA Sheet 2 II.E 3ection IV Evaluation t. Containment N e*"** Not Subject ' 625 To Subenergence I tood Level tw 590 Above Flood Level: Yo.: X No: 00CUMENTATIOM PEFEnENCES W Es 1. FIltL Report F-CIs121s performance qualification tests l of four valve motor operators. g t G D658 OI A
97 i COMPON E$T Limitorque Valve Actuator MO-7068 FUNCTION This ac actuator operates the backup Containment Spray Line Isolation valve which, depending on its position, permits or prevents flow of water from the line - to the secondary Enclosed Spray Nozzle. l DISCUSSION This valve is not required to operate during a break event unless a failure t occurs to MO-7064. As outlined in our submittal to the Director of NRC, dated December 5, 1980, valve MO-7068 is manually opened 15 minutes following primary spray actuation, to provide spray for iodine wasbout from the containment atmosphere. The Limitorque actuator contains a Peerless ac motor with Class B insulation. A Franklin LOCA simulation test documented in Test Report F-C4124, subjected this valve to a 36-hour test which followed the BRP LOCA envelope for this period of time. The test included 24 hours continuous spraying and intermittent sprrty for one hour periods at hours 27, 31 and 35. Relative I humidity was maintained at, or near 100 percent by (1) use of saturated steam to obtain initial temperature anc pressure rise to 2400F, (2) use of fine l mist water over the test specimen, (3) use of saturated steam injections to maintain chamber temperatures. CONCLUSION Raciation and thermal aging qualification testing has not been performed for this type actuator. However, in general most component materials usea in the manufacturing of actuators and motors can withstand a threshold canage limit of at least 4.0 x 106 rads. The effects of thermal and radiation aging are bowever, unknown. Therefore, to meet the guideline requirements, the actuator assembly will be replaced or rebuilt and qualified. JUSTIFICATION FOR INTERIM OPERATICN Interim operation is justified in lieu of qualified equipment based on the following FRC Appendix A Criteria: l 2. Another system is capable of providing the required function of the system with the unqualified equipment. 1 The primary containment spray system (MO-7064) provices sufficient spray flow to limit the temperature rise in containment during a steam line break. Th.4s system is automatically actuated,upon increasing containment pressure 2.2 psig. Rev 3 Doc ID 2240A-211A 3/15/82
l 97a 3. Ite unqualifico equipment will have performed its safety function prior to failure. As ciscussed in the sum. mary paragraph on Containment Atmospheric Concitions, actuation of, Primary Containment Spray (MO-7064) would occur shortly folicwing a steam line break, and before containment environment has significantly degraded (Reference Attachment II). Should the unlikely failure of MO-7064 occur at this time, secondary containment spray (Mo-7068) would be manually actuated immediately. With only limited exposure to the LOCA environment in'this short period, failure prior to actuation is very remote. i 4. The plant can be safely shutdown in the absence of the unqualified equipment. The unique function of the Seconcary Containment Spray is to provide post-accident iodine washdown. Failure of this function would not prevent safe shutdown ano cooldown of the plant. Rev 3 Doc ID 2240A-211A 3/15/82 4 .-- -..~ .m .--U . - - -. ~.
p Krag ATTACm1ENT 3 /
- o UNITED STATES
['g% [*j ,y NUCLEAR REGULATORY COMMIS
- .3 ;
,g WASHINGTON, D. C. 20555 d'" s March 19, 1982 Docket No. 50-155 LS05-82-03-086 Mr.DahidVandeWalle Nuclear Licensing Administrator Consumers Power Company 1945 West Parnall Road Jackson, Michigan 49201
Dear Mr. VandeWalle:
SUBJECT:
BIG ROCK POINT - ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT FOR NUCLEAR POWER PLANTS We have reviewed the information that you provided in your letter dated March 15, 1982 to supplement your 90 day response to the staff's SER and have concluded that you have now provided enough detailed infor-mation to support your justification for interim safe operation. Our review of your submittals is continuing. Sincerely, k E DennisM.Crutchfield,Chidf Operating Reactors Branch #5 Division of Licensing cc: See next page l l Ixt: 4-2. x OS X 03 9Cl x 12. i l}}