ML19339B632
| ML19339B632 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 10/31/1980 |
| From: | Parker W DUKE POWER CO. |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8011070354 | |
| Download: ML19339B632 (16) | |
Text
{{#Wiki_filter:-. DUKE POWER COMPANY Powen Dun.n No 422 Sourn Cucaca Srazzi, CIIARIDTTE N. C. 28242 wimm o. naanen.sn. October 31, 1980 VeCE PetsiCENT m,,,,,,,,.,,,,,o. 57taan Paoouctiom 373-4083 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 M O Attention: Mr. B. J. Youngblood, Chief 3 Licensing Projects Branch No. 1 L.. C-E:.. '2 iiC., ..a L &fG Re: McGuire Nuclear Station, Units 1 and 2 g Docket No. 50-369, 50-370 s 3 m o en w
Dear Mr. Denton:
Please find attached Revision 2 to the document "McGuire Nuclear Station-Evaluation of Cumpliance with NRC Branch Technical Position RSB5-1 on Design Requirements of the Residual Heat Removal System." This information supplements my letters of November 14, 1979 and October 8, 1980. The revision revises our discussion on natural circulation cooldown tests. Please advise if you have further questions on this matter. V ry truly yours, \\ 4 - L$. 2:%,. William O. Parker, Jr. 1.JB :ses Attachment Ok \\ 0 e 801107OM
McGuire Nuclear Station Evaluation of Compliance With NRC Branch Technical Position RSB 5-1 On Design Requirements of the Residual Heat Removal System The following is a discussion of the means by which McGuire Nuclear Station complies with the technical requirements of BTP RSB 5-1, 1. Provide safety grade steam generator dump valves, operators, air and power supplies which meet the single failure criterion. One sa'ety grade steam generator power operated relief valve is provided for eac., of the four steam generators. An air supply to the operators is available during loss of offsite power. The steam generator power operated relief valves can be operated locally to permit plant cooldown. An in plant test shall be conducted to demonstrate local manual operation. lQ212.121 Safety grade remote operators and power supplies are not provided since hot standby can be achieved and maintained using the safety grade steam generator safety valves. See the cold shutdown scenario and single failure evaluation provided belos (Part il - Removal of Residual Heat). 2. Provide the capability to cooldown to cold shutdown in a reasonable amount of time assuming the most limiting single failure and loss of offsite power or show that manual actions inside or outside containment or return to hot standby until the manual actions or maintenance can be performed to correct the failure provides an acceptable alternative. The plant can be maintained in a safe hot standby condition while any necessary manual actions are taken. The plant is capable of being cooled via natural convection and reaching Residual Heat Removal System (RHRS) initiation conditions in approximately 36 hours time including the time Q212.122 regelred to perform any manual actions. See the cold shutdown scenario and single failure evaluation provided below (Part 11 - Removal of Residual Heat). 3. Provide the capability to depressurize the Reactor Coolant System with only safety grade systems assuming a single failure and loss of offsite power or show that manual actions inside or outside containment or remaining at hot standby until manual actions or repairs are complete provides an acceptable alternative. The plant can be maintained in a safe hot standby condition while any required manual actions are taken. See the cold shutdown scenario and single failure evaluation provided below (Part IV - Depressurization). 4. Provide the capability for borating with only safety grade systems assuming a single failure and loss of offsite power or show that manual actions inside or outside containment or remaining at hot standby until manual actions or repairs are completed provides an acceptable alternative. The plant can be maintained in a safe hot standby condition while any required manual actions are taken. See the cold shutdown scenario and single failure evaluation provided below (Part til-Boration and Makeup). Rev. 1 r M
5. Provida ths systca cnd co ponsnt dasign fsstursa nectarary for tha proto-type testing of both the mixing of the added borated water and the cool-down under natural circulation conditions with and without a single failure of a steam generator atmospheric dump valve. These tests and analyses will be used to obtain information on cooldown times and the l corresponding AFW requirements. Part of the low-power testing program at McGuire includes several tests to verify natural circulation. This test program does not include bora-tion or cooldown. Find attached (Attachment A) a comparison of McGuire Unit 1 and 2 and Diablo Canyon Unit i showing that the natural circulation R:v. 2 couldown tests performed at Diablo Canyon will be representative of the Q212.123 natural circulation cooldo.n and boron mixing capability at McGuire. The results of the testing af liiablo Canyon will be reviewed and a natural circulation cooldown-ter sill be performed at McGuire prior to startup following the first refueling if the Diablo Canyon prototype test has not been completed or does not provide satisfactory results. 6. Commit to providing specific procedures for cooling down using natural i circulation and submit a summary of these procedures. l l Specific procedures for cooling down using natural circulation will be prepared and submitted to the NRC by November 19, 1980. A summary of the Q212.121 i procedures is provided in the cold shutdown scenario and single failure. evaluation provided below. l 7. Provide a neismic Category I AFW supply for at least 4 hours at Hot Shut-down plus cooldown to the RRR system cut-in based on the longest time (for only onsite or offsite power and assuming the worst single failure), or show that an adequate alternate seismic Category I source will be available. Sufficient emergency feedwater is available from the Seismic Category I Standby Nuclear Service Water Pond to permit four hours operation at hot standby plus cooldown to RHRS initiation conditions. See the cold shut-down scenario and single failure evaluation provided below (Part II-Removal of Residual Heat). 8. Provide for collection and containment of RHR pressure relief or show that adequate alternative methods of disposing of discharge are available. The RHR relief valves discharge to the prersurizer relief tank, located inside Containment. COLD SHUTDOWN SCENARIO The safe shutdown design basis for McGuire is hot standby. The plant can be maintained in a safe hot standby condition while manual actions are taken to permit achievement of cold shutdown conditions following a safe shutdown earthquake with loss of offsite power. Under such conditions the plant is capable of achieving RHP.S initiation conditions (approximately 3500F, 425 psia) in a reasonable amount of time, including the time required for any manual actions. To achieve and maintain cold shutdown, four key functions must be performed. These are (1) circulation of the reactor coolant (2) removal of residual heat, (3) boration and makeup, and (4) depressurization of RCS. (2) Rev. 2
A 1. Circulation of Reactor Coolant Circulation of the reactor coolant has two stages in a cooldown from hot standby to cold shutdown. The first stage is from hot standby to 350 F. During this stane, circulation of the reactor coolant is provided by natural circulation witt the reactor. core as the heat sourca and steam generators as the heat sink. Steaa release from the steam generators is initially via the steam generator safety valves and occurs automatically as a result of turbine and reactor trip. Steam release for cooldown is via the steam generator power operated relief valves which may be operated manually. The steam generator power operated relief valves are accessible for local operation. Redundant level and pressure indication is provided in the control room for each steam generator. Power for this instrumentation is derived from the 120 VAC vital instrumentation and control power systems. 4 Feedwater to the steam generators is provided by the Auxiliary Fe..Nater System. The AFS is provided with two 100 percent capacity motor driven pu.es and one 200 percent capacity turbine driven pump. Each of the motor driver, aumps sup-plies two steam generators and the turbine driven pump supplies watu to all four steem generators. A seismic Category I source of water for the s75 is available from the Standby Nuclear Service Water Pond which has more than suffi-cient inventory for the longest cooldown time needed with either only onsite or only offsite power available witn an assumed single failure. AFS pump suction switchover to this assured source occurs automatically upon detection of low pump suction pressure. Sufficient safety grade instrumentation will be provided in the control room to monitor AFS operation. The second stage of reactor coolant circulation,is from 350 F to cold shutdown. During this stage, circulation of the reactor coolant is provided by the RHR pumps. 11. Removal of Residual Heat Removal of residual heat also has two stages in a cooldown from hot standby to cold shutdown. The first stage is from hot standby to 350 F. During this stage, the steam generators act as the means o3 heat removal from the Reactor Coolant System (RCS). Initially, steam is released from the steam generators via the steam generator safety valves to mainta in hot standby condi-tions. When the plant operators are ready to begin the cooldown, the steam generator power operated relief valves are opened slightly. As the cooldown proceeds, the operators will occasionally adjust these valves as required to maintain a reasonable cooldown rate. Feedwater makeup to the steam generators is provided from the AuxiliaIy Feedwater System. The Auxiliary Feedwater System has the ability to remove decay heat by providing feedwater to all four steam generators for extended periods of operation. The second stage is from 350 F to cold shutdown. During this stage, the RHRS is brought into operation. The heat exchangers in the RHRS act as the means of heat removal from the RCS. in the RHR heat exchangers, the residual heat is transferred to the Component Cooling System which ultimately transfers the heat to the Nuclear Service Water System. The Component Cooling and the Nuclear Service Water Systems are both designed to Seismic Category I. The RHRs includes two RHR pumps and two RHR heat exchangers. I l 5 e w
Each RHR pump is powered from a dif ferent emergency power train and each RHR heat exchanger is cooled by a different Component Cooling System loop. If any component in one RHR subsystem becomes inoperable, cooldown of the plant is not compromised; however, the time for cooldown would be extended. The status'of the RHRS can be monitored using Class IE instrumentation in the Control Room. If RHRS is unavailable for any reason, cold shutdown may be achieved utilizing alternative methods as discussed in section ll.C of the Single Failure Evalua-Q212.126 - tion. 111. Boration and Makeup Boration is accomplished using portions of the Chemical and Volume Control System (CVCS). Four wt % boric acid from the boric acid tanks is supplied to 4 the suction of the centrifugal charging pumps by the boric acid transfer pumps. ] The centrifugal charging pumps inject the borated water into the RCS via the-i nvemal charging and/or reactor coolant pump seal injection flow paths. Two j boric acid tanks are provided for the plant. They are interconnected so that either tank may be aligned to either unit. Two boric acid transfer pumps are provided for each tank. The boric acid tanks, boric acid transfer pumps, centrifugal charging pumps,'and associated piping are of seismic Category I 3 design. The boric acid transfer pumps and centrifugal charging pumps are powered from emergency power trains. There is sufficient boric acid volume stored in each tank to provide for a cold - I shutdown with the most reactive rod withdrawn. Redundant boric acid tank level indication is provided in the control room. An alternative boration source is the 12 wt % boric acid contained in the boron I injection tank located in the Safety injection System. This source can be used to supplement the boric acid tank to accomplish boration, depending on initial j plant conditions. The contents of the boron injection tank can be delivered to i the RCS by aligning the discharge of the centrifugal charging pumps to this tank j while the suction is aligned to the boric acid tanks. i Makeup, in excess cf that required for boration can be provided from the Refuel-Ing Water Storage Tank (RWST) using centrifugal charging oumps and the same injection flow paths as described for boration. Two motor operated valves, each powered"from different emergency power trains and connected in parallel, would i transfer the suction of the charging pumps to the RWST. RWST level can be monitored using redundant control room instrumentation which has its power derived from the 120 VAC vital instrumentation and control power system. 4 l IV. Depressurization Depressurization of RCS is accomplished using portions of the Chemical and Volume Control System (CVCS). Either four wt. % beric acid or refueling water may be used for depressurization with the flow path being from the centrifugal i ' charging pumps via the auxiliary spray valve to the pressurizer. The centrif-ugal charging pumps of the CVt. Jre of Scismic Category I design and are powered from separate emergency power trains. The pumps can tus operated and monitored j from the control room. Redundant pressurizer level and RCS pressure indication is provided in the control room for monitoring depressurization. Power for this. instrumentation is derived from the 120 VAC vital instrumentation and control power systems. 1 (4) Rev. I t -..-m -e --w-e 9 y v q %w- --m+,-
An alternative method of depressurization consists of discharging reactor coolant from the pressurizer to the pressurizer relief tank via the pressurizer power operated relief valves. V. INSTRUMENTATION Redundan t instrumentation which has its power derived from the 120 VAC vi tal instrumentation and control power system is available in the Control Room to monitor key functions associated with achieving cold shutdown. This instrumen-tation is discussed in FSAR Section 7.5 and includes the following: a. RCS wide range temperature b. RCS wide range. pressure c. Pressurizer water level d. Steam generator narrow range water level e. Steam line pressure f. RWST level g. Containment pressure 4 This instrumentation is sufficient to monitor the key functions associated with cold shutdown and to maintain the RCS within the desired pressure, temperature and inventory relationships. Alternatively, operation of the auxiliary systems that service the RCS can be monitored by the control room operator via remote communication with an operator in the plant. MAINTAINING RCS TEMPERATURE AND PRESSURE DURING C00LDOWN The plant will be maintained in a hot standby condition while the operator eval-uates the initial plant conditions and the availability of equipment and systems (including non-safety grade equipment) that can be used in shutdown. Prior to initiating cooldown, the operator will determine the boration requirements and the method by which the plant will be taken to cold shutdown. In performing the cooldown, the operator will integrate the functions of heat removal, bora-tion and makeup, and depressurization in order to accomplish these functions without letdown from the RCS. Once the plant is cooled to 350 F and depressur-Ized to 425 psia, RHRS operation will be initiated and the RCS will be taken to cold shutdown conditions. Boration, cooldown, and depressurization will be accomplished in a series of short steps arranged to keep RCS temperature and pressure and pressurizer level in the desired relationships. However, to demonstrate that boration and de-pressurization can be done without letdown, a simpler scenario can be used. First the operators integrate the cooldown and boration functions taking advantage of the RCS inventory contraction resulting from the cooldown. Then, the operators use auxiliary spray from the CVCS to depressurize the plant to RHRS initiating conditions. Finally, the RCS is cooled to cold shutdown conditions using the RHRS while makeup with borated water continues as necessary. J (5) Rev. 1
The calculation to demonstrate this capability assumes worst case boration re-outremen:s based on core end of life /pesk xenon conditions and the following RCS Initial conditions following plant trip: RCS Temperature 557 F RCS Pressure 2250 psia Pressurizer Water Volume 450 ft3 Pressurizer Steam Volume 1350 ft The cooldown from 557 F decreases the volume of water in the RCS by approxi-mately 1610 cubic feet assuming that the pressurizer !s not cooled. Makeup for contraction is supplied by 4 wt % boric acid stored in the boric acid tanks 0 at 70 F, A boric acid tank volume of approximately 1450 cubic feet will expcnd to approximately 1610 cubic feet as it is heated to the RCS temperature at 3500F.I The volume of four wt % boric acid at 70 F required for boration to technical specification requirements at 350oF is approximately 1350 cubic feet. Thus the volume required for boration is significantly less than the volume available due to contraction. To calculate if depressurization can be accomplished without letdown and without taking the plant water solid, it was assumed that the pressurizer was initially in the following state: STATE 1 3 Volume, Total, Ft 1800 Volume, Liquid, Ft 450 3 Volume, Steam, Ft 1350 T; = Tsat, F 653 P = Psat, psia 2250 y Quality, X 0.337 It was further assumed that no additional water would be removed from the pressurizer by cooldown contraction. With these assumpstons, and including the effect of heat input from the pressurizer metal, it was determined that spraying in approximately 36,003 lbm of 70 F water would produce the following state: STATE 2 3 Volume, Total, Ft 1800 Volume, liquid, Ft 1180 Volume, Steam, Ft 620 T, = Tsat, F 450 1 P2 = Psat, psia 422.1 (6) Rev. 1
Quality, X 0.0092 i Once depressurized to 425 psla, RHRS operation may be initiated and cooldown can continue to cold shutdown conditions. The cooldown from 350 F to 200 F further decreeses the "olume of water in the RCS by approximately 550 cubic feet assuming that the pressurizer is not cooled. Makeup for contraction is again supplied by 4 wt % boric acid. A boric acid tank volume of approximately 530 cubic feet will expand to approximately 550 cubic feet as it is heated to the RCS temperature of 200 F. The additional volume required for boration at 200 F, to maintain the reactor within the technical specification shutdown requirements, is ro more than 260 cubic feet, the operator having taken full advantage of the previous contraction. Thus, the technical specification requirements for cold i shutdown conditions are satisfied. The results of the calculations described above demonstrate that, based on the assumed initial conditions, boration and depressurization with 4 wt % boric acid can be accomplished without letdown and without taking full credit fc.- the available volume created by the cooldown contraction. However, the operator may elect to borate using the 12 wt % boric acid contents of the boron injection tank as well as 4 wt % boric acid from the boric acid tanks. Should boration without letdown prove impractical due to any combination of plant conditions or equipment failures, letdown can be achieved by discharging RCS inventory via the pressurizer power operated relief valves or the reactor vessel head vent valves. SINGLE FAILURE EVALVATION 1. Circulation of the Reactor Coolant A. From Hot Standby to 350 F (refer to FSAR Figures 5.1-1, 10.3.2-1, and 10.4.7-4) - four reactor coolant loops and four steam generators are provided, any two of which can provide sufficient natural circulation flow to provide adequate core cooling. Even with the most limiting single failure (a steam generator power operated relief valve), three of the reactor coolant loops and steam generators remain available. B. From 350 F to cold shutdown (refer to FSAR Figure 5.5.7-1) - two RHR pumps are provided, either.one of which can provide adequate circula-tion of the reactor coolant. 11. Removal of Residual Heat A. From Hot Standby to 350 F (refer to FSAR Figures 10.3.2-1, 10.4.7-4, and 9.2.2-1) - 1 1. Steam Generator Power Operated Relief Valves - These are air operated valves. Four are provided (one per steam generator), any two of which are sufficient for residual heat removal. in the event of a single failure, three power operated relief valves remain available. In case of air supply failure, these velves fall to the closed position. However, each valve is provided Q212.125 with a handwneel to allow manual control if necessary. The valves are located in the doghouse and are accessible by means of a permanent ladder and scaffolding arrangement. The (7) Rev. 1 mm -~
environment in the doghouse during this cooldown event will not prevent entry for access to the valve. These valves may be reached by an operator in a few minutes, f 2. Auxiliary Feedwater Pumps - Two motor driven pumps and one steam driven pump are provided. In the event of.a single failure, two pumps remain available to provide sufficient i feedwater flow. 3 Auxiliary Feedwater Flow Control Valves CA36, 40, 44, 48, 52, 56, 60, 64 - These are air operated valves. In the event of l a single failure of one flow control valve (which affects flow to one steam generator from either a motor driven pump or the steam driven pump) emergency feed flow can still be j provided to all four steam generators from the other pumps, in case of air supply failure, these valves fall open to a throttled position set to assure adequate flow to the steam Q212.125 ) generators while, at the same time, preventing unacceptable 4 runout of the auxiliary feedwater pumps. As cooldown progresses, flow may be reduced by shutting down auxiliary feedwater pump (s) and manually throttling the control valves with handwheels provided. These valves are located in the auxiliary feedwater pump room and may be reached in a few minutes. The environment 4 in this room during this cooldown event will not prevent entry for access to the valves. Time required for setting each valve is estimat-d to be less than 15 minutes. 4. If the normal non-seismi sources of auxiliary feedwater are not available, automatic re-alignment to the Seismic Category I Standby Nuclear Service Water Pond is provided. Separate and redundant lines provide water to the suction of the AFS pumps. B. From 350 F to 200 F Utilizing RHR System (refer to FSAR Figures 5.5.7-1 9.2.4-1, and 9.2.2-1 through 3) - 4 1. RHR Suction isolation Valves NDIB and N02A - these valves are powered from different emergency power trains. Failure of either power train or of either valve operator could prevent initiation of RHR cooling in the normal manner from the control room. In the event of such a failure, operator action could be taken to open the affected valve manually. The mechanical failure of the disc separating from the stem has been investigated (WCAP-92 and itg probability has been found to be in the range of 10-{7) and to 10" per year The pgobability of an earthquake larger than the OBE is 10-3 to 5x10 > per year. The combined probability of valve stem failure coincident with the earthquake is so low that it need not be considered in the single failure analysis. In the event of such a failure, the plant would remain in a safe hot standby condition with heat removal via the steam generators until an alternative method of cooldown can be established as de-Q212.126 scribed in C below. 2. RHR Pumps A and B - Each pump is powered from a dif ferent emergency power train, in the event of a single failure, either pump can provide sufficient RHR flow. (8) Rev. I l ~..- ~ , ~. r-
3 RHR Heat Exchangers A and B ,if either heat exchanger is unavailable for any reason, the remaining heat exchanger can provide sufficient heat removal capability. 4. RHR Flow Control Valves ND14 and ND29 - These are air operated valves. Upon loss of air these valves would fail in the open position, thus guaranteeing sufficient RHR flow. Q212.125 If a single failure causes one of the valves to fall in a closed or partially closed position,the remaining RHR train can provide sufficient RHR flow. 5. RHR/ SIS Cold Leg Isolation Valves N1173A and Nil 78B - These are parallel, normally open, motor operated valves which are powered from separate emergency power trains. Sufficient RHR cooling flow can be provided through either valve. These valves are also equipped with handwheels for manual operation. 6. Component Cooling System - Two redundant trains are provided, either of which can provide sufficient heat removal capacity via one of the RHR heat exchangers. 7 Nuclear Service Water System - Two redundant trains are pro-vided, either of which can provide sufficient heat removal via one of the Component Cooling System heat exchangers. C. From 350 F to 200 F Utilizing Secondary Plant Systems (refer to FSAR Figures 10.4.7-3 and 10.4.7-4). Cooldown to a main steam temperature of 212 F may be accomplished utilizing natural RCS circulation with auxiliary feedwater to no more than two steam generators and associated power operated relief valves. Coincident with this operation, the remaining steam gener-ators may be prepared for cooidown to a RCS temperature of 200 F utilizing feed and bleed of cold feedwater. As an example, assume steam generators A and B, their associated Q212.126 power operated relief valves, and motor driven auxiliary feedwater pumg A are being used to cooldown to a main steam temperature of 212 F. Concurrently, main feedwater and auxiliary feedwater lines to steam generators C and D may be Isolated, and crossover and drain lines instelled to allow cold lake water to be fed by motor driven auxilUary' feedwater pump B through the main feed nozzles and out the aaxiliary feedwater nozzles which are located well above the tubes. This woujd require filling the generators to a level above the auxiliary feed nozzle. Monitoring of levels in this range may be established utilizing instrumentation provided for steam generator wet layup recirculation. This is but one of several alternative means of achieving cold shutdown. For instance, a less d;sirable but workable method would be to feed and bleed the reactor coolant system utilizing ECCS and the pressurizer power operated relief valves. I (g) Rev. 1 l
Ill. Boration and Makeup (refer to FSAR Figures 9 3.4-1, 2, 3, and 5, and 6.3.2-1) A. Boric Acid Tanks 1 and 2 - Two boric acid tanks are provided with one aligned to each unit. Each tank contains sufficient four wt % boric acid to borate the RCS to cold shutdown with the most reactive rod withdrawn. B. Boric acid Transfer Pumps A and B - Two pumps are aligned to each tank. Each pump is powered from a different emergency power train. In the event of a single failure, either pump can provide sufficient boric acid flow. C. Flow Control Valve NV267A - This is an air operated vaive which fails open on loss of air or power to allow boric acid flow to the Q212.125 suction of the centrifugal charging pumps. MOV NV2658, which is supplied from a separate power train, may be opened to supply boric acid flow directly to the suction header of the centrifugal charging pumps if required. D. Isolation Valves NV17tAand NV175A - These are air operated valves. If either of these valves falls closed, the alternate valve may be opened. If both valves fall closed due to loss of air or power, MOV NV2658 may be opened to supply boric acid flow directly to the suction header of the centrifugal charging pumps. E. Charging Pump Suction isolation Valves NV141A and NV142B - These normally open, motor operated valves are piped in series. If one of these valves closes spuriously, an operator can de-energize the valve operator and reopen the valve with its handwheel. If mechanical failure makes it impossible to open one of these valves, refueling water storage tank isolation valve NV221A or NV222B may be opened to provide makeup flow to the charging pumps. Boration flow may be provided via MOV 265B directly to the suction of the charging pumps. F. Centrifugal Charging Pumps A and B - Pumps A and B are powered from redundant emergency power trains. In the event of a single failure, either pump can provide sufficient boration or makeup flow. G. Norcal Charging Flow Control Valve NV238 - This is an air operated Q212.125 valve which fails open upon loss of air to assure a charging flow
- path, if the valve will not open due to a mechanical problem, a flow path may be established through the Boron injection Tank by opening valves NI4A or N158 and N13A or N1108.
H. Charging Flow Control Valve NV241 - This is an air operated valve which fails open upon loss of air or power to assure a charging flow Q212.125 path. If the valve will not open due to a mechanical problem, a flow path may be established through the Boron injection Tank as explained in G above, l. Charging Line Isolation Valves NV244A and NV245B - if either of these normally open, motor operated valves closes spuriously, an operator may de-energize the valve operator and reopen the valve Rev. 1 (10)
with its handwheel. If this is not possible a flow path can be established through the Boron injection Tank as in G above. J. Reactor Coolant Loop 1 Charging isolation Valve NV138 - This is an air operated valve which fails open upon loss of air or power to assure a charging flow path, it is supplied with Train B emergency power. Loop 4 charging Isolation valve NV16A, which Q212.125 .also falls open upon loss of air or power, may also be opened to provide a charging flow path. NV16A is supplied with Train A emergency power. 4 K. Boron injection Tank isolation Valves NihA and N158 - Each valve is powered from a different emergency power train; only one of thesc normally closed, motor operated valves needs to be opened to provide an alternate path and source for boration. L. Baron injection Tank isolation Valves Nf 3A and N110B - Each valve is powered from a different emergency power train; only one of these normally closed, motor operated valves needs to be opened to provide an alternate path and source for boration. M. Refueling Water Storage Tank isolation Valves NV221A and NV2228 - Each valve is powered from a different emergency power train. Only one of these normally closed motor operated valves needs to be opened to provide an alternate makeup flow path from the RWST to the centrifugal charging pumps. IV. Depressurization (refer to FSAR Figure 9.3.4-1) A. Auxiliary Spray Valve NV21A - This is an air operated valve which falls closed on loss of air or power. In this case, NV21A may be opened by using a portable compressed air or nitrogen bottle. This valve is located in the pipe chase behind the crane wall inside the Containment Building and is readily accessible. Estimated amount Q212.l*5 4 of time to perform this oper ation is one hour. Temperatures and pressures in this area for this cooldown mode will be in the normal range (<l20 F, 0 psig). Radioactivity at this location will not hinder the operation. If NV21 A is stuck closed as a result of mechanical failure, the redundant Seismic Category 1 pressurizer power operated relief valves may be used to depressurize the RCS by discharging to the pressurizer relief tank as described in item C below. B. Charging Valves NV16A and NVl3B - These air operated valves fail open upon loss of air or power. In this case, NV16A and NV138 may be closed by using portable compressed air or nitrogen bottles. i Both of these valves are located in the pipe chase behind the crane wall inside the Containment Building and are readily accessible. Q212.125 Estimated amount of time to perform this operation is less than one hour.. Temperatures and pressures in this area for this cooldown mode will be in the normal range (< 120 F, 0 psig). Radioactivi ty at this location will not hinder the operation. Rev. 1 (;g .,s -n p -p
If NV16A or NVl3B is stuck open as a result of mechanical failure, the redundant Seismic Category 1 pressurizer power operated relief Q212.125 valves may be used to depressurize the RCS by discharging to the pressurizer relief tank as described in item C below. C. Pressurizer Power Operated Relief Valves, NC32B, NC34A, NC36B - These are air operated valves which fall to the closed position upon loss of air or power. As Indicated above, these valves may be used to depressurize the RCS. The operator may open them from the control room, thus discharging steam from the pressurizer to the pressurizer relief tank. If the normal air supply is not aval-able to actuate these valves, a supply of nitrogen may be made Q212.125 available to either NC34A or NC32B from cold leg safety injection accumulators l A or IB, respectively, which provide the seismically qualified nitrogen supply necessary for low RCS temperature over-pressure protection (see FSAR Section 5.2.2.3). This is accomplished j by placing Jumber cables across the low temperature permissive con-tacts in the area termination cabinets for either block valve N1430A or N14318. The valve may then be opened from the control room, thus providing nitrogen for actuation of the associated relief valve. 1 D. RHR Suction isolation Valve NDIB and ND2A - The RHR suction isolation valves are qualified for the steam line break environment. Therefore, they are qualified for the less severe environment that would result if, as described in C above the RCS is depressurized by discharging the pressurizer to the pressurizer relief tank. V. Instrumentation Sufficient instrumentation is provided in the control room to monitor key functions. In the event of a single failure, the operator can make comparisons between duplicate information channels or between functionally related channels in order to identify the particular malfunction. Refer to FSAR Section 7.5 for applicable details. k i i (y g) Rev. 1 +-- w w
ATTACHMENT A McGuire Nuclear Station Evaluation of Compliance With NRC Branch Technical Position RSB 5-1 On Design Requirements of the Residual Heat Removal System Comparison of McGuire to Diablo Canyon McGuire Unit I and Unit 2 and Diablo Canyon Unit I have been compared in detail to ascertain any differences setween the two plants that could potentially affect natural circulacion flow and attendant boron mixing. The general configuration of the piping and components in each reactor coolant loop is the same in both McGuire and Diablo Canyon. Both plants have Model 93A reactor coolant pumps. McGuire Unit I has Model D2 6ceam generators (McGuire Unit 2 has Model D3) and Diablo Canyon has Model 51 steam generators. The elevation head and flow resistances represented by these components and the system piping is similar. To compare the natural circulation capabilities of McGuire and Diablo Canyon, the hydraulic resistance coefficients were compared. The hydraulic resistance coefficients applicable to normal flow conditions are as follows: Diablo Canyon Unit 1 McGuire Units 1&2 -10 Reactor Core & Internals 8.0 x 10-10ft/(loop gpm)2 6.91 x 10 Reactor Nozzles 36.8 27.60 RCS Piping 24.0 24.00 Steam Generator 114.0 110.66 TOTAL LOOP 182.8 169.17 ,(182.8)V2 Flow Ratio McGuire 1.04 = Diablo Canyon 169.2 The general arrangement of the reactor core and internals is the same in McGuire and Diablo Canyon. The coefficients indicated represent the resistance seen by the flow in one loop. The reactor vessel outlet nozzle configuration for both plants is the same. The radius of curvature between the vessel inle' nozzle and downcomer section of the vessel on the two plants is different. Based on 1/7 scale model testing performed by Westinghouse and other literature, the radius on the vessel nozzle / vessel downcomer juncture influences the hydraulic resistance of the flow turning from the nozzle to the downcomer. The Diablo Canyon vessel inlet nozzle radius is significantly smaller than that of McGuire, as reflected by the higher coefficient for Diablo Canyon. The coefficient of resistance for the RCS piping for both plants is the same.
Fluid may also be exchanged between the upper plenum region (i.e., the portion of the reactor between the upper core plate and the upper support plate) and the upper head region via the guide tubes and support columns. Guide tubes and support columns are dispersed in the upper plenum region from the center to the periphery. Because of the non-uniform pressure distribution at the upper core plate elevation and the flow distribution in the upper plenum region, the pres-sure in the support columns and guide tubes varies from location to location. These support column and guide tube pressure variations create the potential for flow to either enter or exit the upper head region via the support columns or guide tubes. To ascertain any difference between the upper head cooling capabilities between Diablo Canyon and McGuire, a comparison of the hydraulic resistance of the upper head regions was made. These flow paths were considered in parallel to obtain the following results. McGuire Units 1 & 2 Diablo Canyon Unit 1 Flow Area (ft ) 2.74 0.77 Loss Coefficient 1.96 1.51 Overall Hjdraulic Resistance 0.26 2.57 (ft) Relative Head Region Flow Rate 3.14 1.00 As indicated above the effective hydraulic resistance to flow in McGuire is only 10% of that in Diablo Canyon. Assuming that the same pressure differential existed in both plants the McGuire head flow rate would be three times the Diablo Canyon flow. Thus, the upper head cooling capability at McGuire would be no worse and would likely be better than demonstrated by the Diablo Canyon natural circulation cooldown test. In conclusion, the results of the natural circulation cooldown tests performed at Diablo Canyon will be representative of the natural circulation and boron mixing capability of McGuire, and the results of these tests will be reviewed. A natural circulation cooldown test will be performed at McGuire prior to start-up following the first McGuire refueling if the Diablo Canyon prototype test has not been completed or does not provide satisfactory results. Details of the specific steam generator units were also compared to ascertain any variation (e.g., primary volume, tube height, tube diameter) that could affect natural circulation capability by changing the effective elevation of the heat sink or the hydraulic resistance seen by the primary coolant. It was concluded that there are no differences in the design of the steam generators in these ple.nts that would significantly affect the natural circulation characteristics. As indicated, the difference between the total resistance coefficients for the two plants is insignificant. It is expected that the relative effect of the coefficients would be the same under natural circulation conditions resulting in a natural circulation loop flow rate for McGuire within 4 percent of that for Diablo Canyon.
The coefficients provided reflect the flow rate and associated heat removal capability of an individual loop in the plant. The comparison, therefore, does not take into consideration the number of loops available nor the core heat to be removed. An evaluation of the McGuire Steam Relief and Auxiliary Feedwater Systems has been performed to demonstrate that cooling can be pro-vided via three steam generators following the most limiting single active failure, i.e., the failure of an atmospheric relief valve. Loop natural circulation flow is dependent on reactor core decay heat which is a function of time based on core power operating history. Under natural circulation flow conditions, flow into the upper head area will constitute only a small percentage of the total core natural circulation flow and there-fore will not result in an unacceptable thermal / hydraulic impedance to the natural circulation flow required to cool the core. For typical 4-loop plants (including McGuire and Diablo Canyon), there are three potential flow paths by which flow crosses the upper head region boundary in a reactor. These three paths are the head cooling spray nozzles, the support columns (l) and the guide tubes. The head cooling spray nozzle is a flow path between the downcomer region and the upper head region. The temperature of the fluid which enters the head via this path corresponds to the cold leg value (i.e., Tcold)- i 1. Three for a UHI plant: support columns are not a potential path for the non-UHI plant. I --}}