PY-CEI-NRR-2362, Application for Amend to License NPF-58,proposing Change in Plant USAR by Incorporating leak-off Line in RHR Sys, Which Will Significantly Reduce Collective Dose to Plant Operations Personnel.Annotated USAR Pages,Encl
| ML20204F385 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 03/17/1999 |
| From: | Myers L CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| PY-CEI-NRR-2362, NUDOCS 9903250278 | |
| Download: ML20204F385 (31) | |
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{{#Wiki_filter:S -s Perty Nuclear Power Plant P ox 7 m Perry, Ohio 44081
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Lew W Myer'd '.' 440-280-5915 Vice Presicient Fax:440-280 8029 l March 17,1999 PY-CEl/NRR-2362L United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 . l Perry Nuclear Power Plant Docket No. 50-440 License Amendment Request: Modification to the Residual Heat Removal System l Ladies and Gentlemen: Pursuant to 10CFR50.59 and 10CFR50.90, Nuclear Regulatory Commission review and approval is requested on a proposed modification that changes the Perry Nuclear Power Plant Unit 1 as described in the Updated Safety Analysis Report (USAR). The change incorporates a leak-off line in the Psesidual Heat Removal system. The leak-off line is designed to el;.ninate an operator work around, which will significantly reduce the collective dose to plant operations personnel. Attachment 1 provides a Summary, Description of the Proposed Modification, a Safety Analysis, and an Environmental Consideration. Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides a copy of the annotated USAR pages. There are no regulatory commitments contained in this letter or its attachments. 4 If you havo questions or require additional information, please contact Mr. Henry L. Hegrat, Manager - Regulatory Affairs, at (440) 280-5606. Very truly yours, x c)mp Attachments cc: NRC Project Manager NRC Resident inspector b)I j NRC Region ill State of Ohio 9903250278 990317 ' PDR ADOCK 05000440 P PM ;
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l 1 I, Lew W. Myers, being duly sworn state that (1) I am Vice President - Perry, of the ! FirstEnergy Nuclear Operating Company, (2) I am duly authorized to exccute and file { this certification as the duly authorized agent for The Cleveland Electric Illuminating ) Company, Toledo Edison Company, Duquesne Light Company, Ohio Edison Company, and Pennsylvania Power Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief. EWN/u p 'ers I Sworn to and subscribed before me, the /7 day of ideud. ,/9f/ W
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Attichm:nt 1 PY-CEl/NRR-2362L Page 1 of 15
SUMMARY
In accordance with 10CFR50.59 and 10CFR50.90, Nuclear Regulatory Commission (NRC) review and approvalis requested on a proposed modification to the Perry Nuclear Power Plant (PNPP) Unit 1 as described in the Updated Safety Analysis report (USAR). The change incorporates a leak-off line in the Residual Heat Removal (RHR) system. The leak-off line is designed to eliminate an operator work around, which will significantly reduce the collective dose to plant operators [ collective annual operator dose received due to the wo'K around is approximately one (1) rem]. This modification is being submitted for NRC review and approval based upon a completed 10CFR50.59 Safety Evaluation which determined that a portion of the proposed modification, specifically the re-designation of valve 1E12-F073B (outboard containment i isolation valve for the RHR Heat Exchanger "B" vent line penetration) from a manual operated, "normally closed' valve to an automatic operated, "normally open" valve had resulted in a slight " increase in the probability of a malfunction of equipment important to safety previously evaluated in the safety analysis report." Although this proposed modification is considered to provide a significant safety benefit and improve the reliability of the RHR system,10CFR50.59 requirer submittalin accordance with 10CFR50.90 due to the above determination. DESCRIPTION OF THE r3 ROPOSED MODIFICATION 1 BACKGROUND The Shutdown Cooling (SDC) feature of the RHR system is used during shutdown operations to control reactor pressure vessel and reactor coolant piping temperature. The ! SDC header is isolated from the Reactor Recirculation system by three Pressure Isolation Valves (PlVs) 1E12-F008,1E12-F009, and 1E12-F550 (refer to Figure 1). During normal ' plant operations the PlVs are closed. Normal design leakage from 1E12-F008 and l 1E12-F009 causes the isolated, water-filled, SDC header to heat up and pressurize. When j the pressure in this portion of the RHR piping reaches approximately 180 psig, an alarm j sounds in the control room. The response to this alarm is to vent the SDC header in i accordance with approved plant operating instructions. If the pressure in the pipe reaches approximately 185 psig, an over-pressure protection relief valve,1E12-F005, will open to , relieve header pressure. Since the fluid in the SDC header is at elevated temperatures and pressure, it may flash to steam as it exits the relief valve. In order to avoid excessive wear on the relief valve, and to reduce dose to the plant operators associated with the venting operations, plant operators periodically drain the water from the SDC header. This activity has extended the time period between successive venting operations and successive lifts of the relief valve. To provide a long-term resolution for this issue, an evaluation was performed with a variety l of solutions postulated. The modification proposed by this license amendment request was selected as the best solution since it will eliminate the need to perform both the venting and draining operations, eliminating an operator work around and reducing operator dose. l J
Attachm;nt 1 PY-CEl/NRR-2362L Page 2 of 15 PROPOSED MODIFICATION The proposed modification (Figure 1) adds a %-inch leak-off line between the 20-inch SDC header and the 1-inch RHR "B" Heat Exchanger (HX) vent line. This creates a flow path from the SDC header to the suppression pool via two valves 1E12-F073B and 1E12-F5588. Valve 1E12-F073B is the outboard containment isolation valve for the RHR (
"B" HX vent line. This flow path will be aligned during normal plant operation to relieve SDC header pressure.
The cross-connection between the SDC header and RHR HX vent line will be installed on an existing RHR flush line near high point vent valve 1E12-F581. The piping will then run approximately ten feet, and terminate in the RHR "B" HX vent line between Motor Operated Valves (MOVs) 1E12-F073B and 1E12-F074B. The flow path continues past 1E12-F073B and into the containment. Once in containment, the flow goes through 1E12-F5588 and terminates beneath the suppression pool water level in the drywell weir. A new %-inch globe valve 1E12-F620 will be installed in the leak-off line near 1E12-F581 to provide manual isolation capability for the leak-off line. The leak-off line piping and maintenance isolation valve are located within the RHR "B" HX room in the Auxiliary Building. MOV 1E12-F073B is a remote manual valve that is currently normally closed during plant operation. The valve requires neither automatic control functions nor automatic isolation functions. In order to align the leak-off line for SDC pressure control,1E12-F073B will need to remain open during normal plant operations. The valve's Loss-Of-Coolant-Accident (LOCA) position will remain as the closed position. To facilitate the closure of the valve during a LOCA, the valve will be upgraded to include an automatic LOCA isolation signal. The valve will automatically close when the following isolation signals are received: reactor vessel low water level (Level 3), high drywell pressure, or a RHR system line break (high ambient temperature or high differential temperature sensed in the RHR "B" room). To convert 1E12-F073B to an automatic valve, modifications will be made at the MOV, the associated Motor Control Center (MCC), and in the control room. The closure signal will be obtained from spare contacts on an existing RHR LOCA relay. This relay is normally energized and will de-energize (relay de-energization will initiate the desired actions) upon receipt of either reactor vessel low water level (Level 3), high drywell pressure, or high ambient or high differential temperature in the RHR "B" area from the Leak Detection system. The conversion of the valve from manual to automatic impacts the diesel generator loading. The valve is currently listed in the diesel generator loading calculation as not being required following a LOCA. As a valve designed to automatically close post-LOCA, the motor load will be required to be added to the diesel generator loading following a LOCA. The results of the proposed modification will add approximately 0.22 kW to the diesel generator post-LOCA, which does not adversely affect the diesel generator. As a part of this modification, administrative controls will be established to limit the total allowable Pressure Isolation Valve (PlV) leakage from the RHR containment penetration P421 (1E12-F008 serving as the outboard barrier, and 1E12-F009 and 1E12-F550 trying as the inboard barrier) to less than 0.30 gpm. The 0.30 gpm (total) penetration leaLge value is significantly below the Technical Specification 3.4.6 leakage limit of less than
i Attachmint i PY-CEl/NRR-2362L Page 3 of 15
- 5.0 gpm (0.50 gpm per nominalinch of valve size up to a maximum of 5.0 gpm, leakage from this penetration is less than 5.0 gpm since the valve is 20" in size).
SAFETY ANALYSIS RHR SYSTEM CONSIDERATIONS I The RHR system has six modes of operation: Shutdown Cooling (SDC) Mode, Low Pressure injection (LPCI) Mode, Suppression Pool Cooling Mode, Containment Spray Cooling Mode, Fuel Pool Supplemental Cooling Mode and Containment Flood Mode. A .. recent change to the PNPP Updated Safety Analysis Report (USAR) has deleted the L Reactor Steam Condensing Mode as a mode of the RHR system operation. With respect to this proposed modification, the key RHR operational mode is the SDC Mode. The design basis of the Shutdown Cooling Mode is to have the capability to remove decay and sensible heat from the reactor primary system, reducing the reactor outlet temperature to 125 F within 20 hours after the control rods heve been inserted. The SDC portion of the RHR system was designed and evaluated against the acceptance criteria of NUREG-0800, Section 5.4.7, " Residual Heat Removal (RHR) System", and Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System", which state the SDC portion of the RHR system must meet the General Design Criteria (GDC) for safety-related systems (i.e., GDC 1, GDC 2, GDC 3, GDC 4, GDC 19, and GDC 34). The Shutdown Cooling Mode, which is a safety-related function, is designed for the most limiting single failure, which is described in the USAR as the loss of an entire RHR heat exchanger loop concurrent with the loss of one of the Reactor Recirculation system loop suction valves. In other words, one loop of RHR would be inoperable and the SDC header would also be inoperable. In this event, the reactor primary system would be cooled down using the Alternate Shutdown Cooling system. The Alternate Shutdown Cooling system transports the heat from the reactor vessel via the safety relief valves to the suppression pool. The suppression pool is then cooled using the operable loop of the RHR system in the Suppression Pool Cooling Mode. A key feature to the SDC configuration and to the proposed modification is the SDC header. The SDC header provides the suction path from the reactor vessel to support operation of the Shutdown Cooling Mode. The SDC header is a common component that can serve either or both of the RHR heat exchanger loops. Note, the SDC header only suppo:ts SDC operation and not the other RHR functional modes.
. SDC HEADER PIPING IMPACT i The SDC header was designed to comply with GDC 1. As such, the header is classified as ASME Section ill, Class 2. All pressure retaining piping modified or added as a result of the proposed modification will conform to ASME Section Ill, Division 1, Subsection NC (Class 2 pipina). The proposed piping was analyzed to ensure that the stress limits for normal, ,
I upset, emergency, and faulted conditions were in accordance with the applicable portions of ' the ASME code associated with mechanical system components. The results of the analyses demonstrated that the resulting stresses are within the RHR system design bases.
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Attachment 1 PY-CEl/NRR-2362L Page 4 of 15
- SDC OPERATIONAL IMPACT i
As shown on Figure 1, the new %-inch leak-off line interconnects the 20-inch SDC header with the suppression pool via the RHR "B" HX vent line. This line will be open during normal plant operation. E aring the SDC Mode of operation, valve 1E12-F073B is required to be closed to isolate the leak-off line. The leak-off line does not provide any flow paths for this mode of RHR system operation. 1E12-F073B will be closed by the control room operators as a part of the procedure to initiate the SDC Mode of operation. The ability to remote manually operate 1E12-F0738 from the control room satisfies the RHR system design requirements that the SDC function be capable of being operated from the control room. The SDC Mode of RHR system operation is manually initiated. In accordance with the plant operationalinstructions, the RHR system must be filled, vented, flushed and warmed-up prior to initiating this mode of operation. To perform these evolutions, various valve manipulations must be performed, both in the control room and in the plant. These valve manipulations include the opening and closing of valves 1E12-F073A(B) and 1E12-F074A(B). The installation of the proposed modification will not change the need to manipulate various RHR system valves. The addition of the leak-off line raises two issues relating to the operation of the SDC Mode of operation: (1) the failure to close 1E12-F073B during the initiation of SDC, and (2) the consequences of +he inadvertent opening of 1E12-F0738 valve while in the SDC Mode. Either of these failures during the SDC Mode of operation would result in the diversion of a portion of the reactor coolant to the suppression pool. With respect to the first issue, procedural controls will be established to close and verify closed 1E12-F0738 prior to initiating the SDC Mode of operation. If the valve is not closed, the SDC Mode will not be permitted to be initiated. Hence, reactor coolant will not be diverted. With respect to the second concern, procedural controls regarding SDC operation would ensure that 1E12-F073B is not opened as a result of inadvertent operator action. If the valve inadvertently opened as a result of some type of equipment failure, this would permit flow diversion. However, this event could happen regardless of whether the modification was implemented or not. Valve indication for 1E12-F073B is located in the control room. This indication would alert control room operators to potential valve mispositioning problems. The automatic isolation function added by the modification will automatically c!cse 1E12-F073B when Level 3 is reached. The control room operator can terminate the event by either closing 1E12-F073B from the control room, by having a plant operator manually shut 1E12-F073B, or by having a plant operator manually shut ! 1E12-F620 (the leak-off line isolation valve installed as part of the proposed modification). It should be noted that a plant operator would be in the vicinity of the valves since the operator is required to locally manipulate RHR system valves as part of the initiation process of the SDC Mode of operation. ALTERNATE SHUTDOWN COOLING IMPACT The Alternate Shutdown Cooling flow path injects suppression pool water via either the RHR system Low Pressure Coolant injection (LPCI) "C" or the Low Pressure Core Spray (LPCS) system into the reactor vessel. The water then returns to the suppression pool via the
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l Attachment 1 PY-CEl/NRR-2362L Page 5 of 15
- safety relief valves. The suppression pool water is cooled by the Suppression Pool Cooling Mode of RHR.
The proposed leak-off line is installed in the RHR system "B" loop. Any failure of the RHR "B" loop will not affect the "C" loop since the two loops are mechanically and electrically independent. The leak-off line is not connected to either the LPCS system or the safety relief valves. Therefore, failure of the leak-off line or failure of 1E12-F0738 to shut will not affect the Alternate Shutdown Cooling path. USAR Section 15.2.9, " Failure of RHR Shutdown Cooling", analyzes a loss of SDC. The analysis is based upon a worst case scenario in which neither SDC loop is available. The event is mitigated by the initiation of the Altemate Shutdown Cooling Mode of operation. As
' stated above, the failure of the leak-off line or the failure of 1E12-F0738 to close does not affect the availability of the Altemate Shutdown Cooling Mode. Hence, the USAR event will not be impacted.
PRESSURIZATION OF THE SDC HEADER DUE TO VALVE LEAKAGE HHR system suction isolation valves 1E12-F008 and 1E12-F009 were originally designed and procured with an allowable leakage rate of approximately 0.05 gph. Reactor coolant at normal operating temperature and pressure at this leakage rate when introduced into the SDC header (header is filled with relatively cool water at low pressure) will pressurize almost immediately. The SDC header is protected from over-pressurization by pressure transmitter 1E12-N057 and by relief valve 1E12-F005. The pressure transmitter provides an alarm (early warning) function. The relief valve provides the code over-pressure protection. 1 During normal plant operation, there is a possibility that the 1E12-F0738 could inadvertently I close, e.g., loss of a RPS bus or an operator error, if the 1E12-F0738 valve were closed during normal operation, the system configuration would be as it is currently. The pressure transmitter will alarm in the control room and 1E12-F005 would provide over-pressure protection if needed. FAILURE TO CLOSE 1E12-F074B 1E12-F074B could be open, during various plant conditions, to vent the RHR heat exchanger. To perform the venting operation,1E12-F0738 must also be open. Hence, an ! open path to the suppression pool would exist. Therefore, the effects of the failure to close i 1 E12-F074B are described in the following paragraphs. l In the current configuration, if a postulated accident event occurred while venting the RHR heat exchanger, the failure to close manual valvo 1E12-F074B would divert a small amount of the LPCI or containment spray flow from the RHR heat exchanger to the suppression pool. In the proposed configuration, however, the flow to the suppression pool through the ! 1-inch vent line in which 1E12-F074B is located would be terminated by the automatic closure of 1E12-F073B. Flow would, however, continue to the suppression pool through the
% inch leak-off line via the SDC he .ar relief valve 1E12-F005 (LPCI mode pressure during accident conditions would exceed the relief valve setpoint). The flow to the suppression pool assuming 1E12-F074B is open in the proposed configuration would be less than in the current configuration, i.e., flow through a %" line will be less than flow through a 1" line
i l Attachm:nt 1 PY-CEl/NRR-2362L Page 6 of 15 The failure to close 1E12-F074B during SDC operation could not create conditions that could cause an inadvertent reactor level drain down to the suppression pool via the SDC relief valve. With the RHR system Shutdown Cooling interlock set at 135 psig (if pressure exceeds the interlock value, SDC operation cannot be initiated) and the set point for the 1E12-F005 (relief valve) set at 185 psig, pressure in the SDC header and leak-off line would never the reach the relief setpoint. Hence, there will be no flow to the suppression pool via j the relief valve. I The failure to close 1E12-F074B while the RHR system is in stand-by readiness could not affect the operational status of the system. While in stand-by readiness, the RHR " water leg" pump maintains the system full of water. With the proposed modification installed, if 1E12-F0748 was left open, the water leg pump would still maintain the RHR system full of water since the head of the water leg pump is below the SDC relief valve setpoint (185 psig). Hence, the lines will remain pressurized and full. 1 IMPACT UPON THE OPERABILITY OF 1E12-F558B ) The proposed modification evaluated the potential impacts upon the inboard containment isolation valve,1E12-F558B, due the "open/close" cycling dur'.ng leak-off line operation. 1E12-F558B is a 1" spring-loaded lift check valve. The valve is expected to cycle due to the minimal leak-off flow of 0.30 gpm and the AP created by the flow of approximately 5 psid which is required to open the check valve. The tha mal / hydraulic analysis (described later within this attachment) in conjunction with admin:stratively controlled leakage limits shows that the fluid flow velocity will be very low and that the fluid will not flash as it passes though the check valve. Therefore, the cycling caused by the proposed modification will not adversely affect the check valve. CONTAINMENT ISOLATION SYSTEM IMPACT I The design basis of the containment isolation system is to allow the passage of fluids (during both normal and emergency conditions) through the containment while preventing or limiting the escape of fission products that may result from postulated accidents so that the radiological consequences do not exceed 10 CFR 100 limits (USAR Section 6.2.4), USAR Section 6.2.4," Containment isolation System", identifies that the containment isolation system design includes compliance with the requirements of GDC 1,2,4,16,54,55,56, , and 57, 1 1 The proposed modification routes the RHR system valve seat leakage from 1E12-F008 and l 1E12-F009 to the suppression pool by uti;izing the existing vent line for the PHR "B" heat } exchanger. The existing containment isolation provisions for the RHR heat exchanger vent line penetration, P431, includes an inboard check valve,1E12-F558B, and ar, outboard motor operated valve,1E12-F073B. In addition to the two mentioned containment isolation valves, this line also includes an additional RHR isolation valve,1E12-F0748, that is not a l containment isolation valve. This modification routes the leakage to the suppression pool by ' installing a connection to the SDC header between two existing "normally closeo" valves, 1E12-F073B and 1E12-F0748, and by changing the 1E12-F073B valve from a manually operated, "normally closed" to an automatically operated, "normally open" valve. The 1E12-F074B vaive will remain a normally closed remote manually operated valve.
Attachment i PY CEl/NRR-2362L 1 Page 7 of 15 I
- The original configuration of this penetration was evaluated against the requirements of J l
GDC-56. GDC 56 requires a penetration to satisfy one of the following configtvations: l One locked closed isolation valve inside and ono locked closed isniation .! valve outside of containment, or-One automatic isolation valve inside and one locked closed isolation vaNo outside of containment, or One locked closed isolation valve inside and one automatic isolation valve outside of containment (a simple check valve cannot be used as the automatic isolation valve outside containment), or One automatic isolation valve inside and one automatic isolation valve outside of containment (a simple check valve cannot be used as the automatic isolation valve outside containment), or I Containment isolation provisions are acceptable on some other defined basis. In USAR Section 6.2.4.2.2.2.a.3, the RHR heat exchanger vent line penetration (P431) was I evaluated along with the general category of suppression pool penetrations that do not ) explicitly meet GDC-56. The justification in the USAR for this par 1icular penetration states ; that "mo normally closed remote-manually actuated, motor operated valves outside containment, and a check valve, located between containment and the drywell, provide isolation". Acceptance of this configuration was not based upon the use of " locked" closed valves. Furthermore, the existing configuration considered the presence of the two motor 1 operated valves even though one of them,1E12-F0748, is not a containment isolation valve I (refer to USAR Table 6.2-32, Containment isolation Valve Summary). As evident in the licensing basis, the valves were considered "normally closed," however, the licensing basis does not identify any requirement to " maintain" the valves closed at all times during plant operation nor does the basis identify an acceptable frequency of manual operation. In the l current configuration, the valves are remote manually opened, during plant operation, as i needed to periodically relieve pressure in the RHR *B" heat exchanger or to warm-up the i RHR heat exchanger prior to the initistion of the Shutdown Cooling Mode. The current l licensing basis allows this operation, and hence recognized that a periodically opened valve with reliance on operator action to close the valve, represented an acceptable level of risk. The proposed modification converts penetration P431 to a configuration that fully meets the requirements of GDC 56. In the proposed configuration the penetration now includas one automatic isciation valve inside containment (the spring loaded lift : heck valve, 1E12-F5588) and one automatic isolation valve outside containme nt (1E12-F0738). Even though the proposed modification changes the position of 1E12-F073B from "normally closed" to "open", the overall reliability of the penetration will be improved. The improved reliability is based upon the following; 1.) the vulnerability to operator error from the successful nerformance of the safety-function (valve closure) has been eliminated (by the incorporati n of 'he automatic c!osure), and 2.) the modification reduces the cycling and wear on 1L12-F005 (overpressure protection relief valve. OTHER CONTAINMENT SYSTEM CONSIDERATIONS i Leakage through the RHR vent line penetration, P431, does not directly communicate with the atmosphere outside of containment. The penetration connects to the RHR "B" heat exchanger through the 1E12-F0748 valve and the SDC header through the new '%" leak off m
Attachment 1 PY-CEl/NRR-2362L Page 8 of 15
- connection. Both of these paths are to a closed, safety-related fluid system. The fluid leakage from the RHR system is controlled as a part of the program required in accordance with Technical Specification 5.5.2, ." Primary Coolant Sources Outside Containment."
The proposed modification provides a signal to automatically close valve 1E12-F0738 as required for containment isolation. Vcive closure signals will be either high drywell pressure, low water level (Level 3),' or high temperature or differential temperature in the RHR equipment area as determined by the Leak Detection System. The isolation. of this penetration with the high drywell pressure and low water level (Level 3) is consistent with
- the PNPP methodology described in USAR Sections 6.2.4 and 7.3.
The current design imposes a severe operating service on the existing SDC header relief valve,1E12-F005. As stated above,1E12-F005 is a containment isolation valve. During pressurization of the SDC header, this valve starts to cycle to relieve the pressure in the SDC header. The cycling of this valve continues until the operators take manual actions to vent the SDC header. The addition of the vent line will significantly reduce the cycling of the relief valve. This clearly increases the reliability of the relief valve and reduces the probability of malfunction of this containment isolation valve. The conversion of a manually operated, "normally closed" valve to an automatically operated, "normally open" valve (1E12-F073B) raises questions regarding the probability of the component to not perform its safety function (to close). The existing licensing basis does not require this mctor operated valve be " locked" closed or " maintained" closed at all times during power operation, and therefore, it can be concluded that the existing configuration includes a probability of failure to some degree. The most probable failure would be failure of the operator to re-close the valve after usage, if the valve was inadvertently left open after the performance of some activity and an accident occurred - (accident position of the valve is closed), it could be postulated that due to +.he stress of and the demands placed upon the operator during the event, the valve would not be closed in a i timely manner. From a reliability perspective, it would then be apparent that closing an open automatic valve is more reliable than closing an open manual valve. In the absence of probability analysis in the design and licensing basis, it is assumed that compliance with the , applicabla GDC affords an acceptable level of reliability. It can also be assumed that the ! current licensing basis, as an exception to the GDCs, provides a lesser degree of reliability. Therefore, it is concluded that the proposed GDC compliant penetration is inherently an improvement in reliability. The conversion of a manually operated, "normally closed" valve to an automatically operated open valve does not affect the postulated radiological consequences of an assumed accident from a closure time perspective. The radiological consequence analyses ) throughout the industry do not consider the release of activity during the closure of the Containment Isolation Valves (CIVs) (purge valves being an exception). The ClV closure times are based upon supporting the core cooling function cad the effectiveness of the reactor vessel blowdown to the suppression pool. This would not be an issue for this penetration. ELECTRICAL AND CONTROL SYSTEMS IMPACT The proposed modification converts Motor Operated Valve (MOV) 1E12-F073B from a manual operated, normally closed containment isolation valve to an automatic operated, normally open containment isolation valve. E6ctrical modifications will be made at the
Attachment 1 PY-CEt/NRR-2362L Page 9 of 15
.MOV, the associated Motor Control Center (MCC), and the associated control room control panels. The MOV closure signal will be obtained from t pare contacts of existing RHR LOCA relay 1821H-K598. This relay is nonnally energized and will drop out upon either high drywell pressure, low water level (Level 3), or high temperature or differential temperature in the RHR equipment area as determined by the Leak Detection System.
The maximum allowable MOV stroke time as specified in plant surveillance instructions wi!! not be altered by the proposed modification. I Since 1E12-F073B will now have an automatic closing function, a new switch legend will be i installed with the center position being AUTO (for automatic) to prevent any Human Factors concem. I The conversion of the MOV from manual to automatic imoacts the diesel generator loading. The vcive is currently listed in the diesel generator loading calculation as not being required following a LOCA. As an automatic valve, the motor load will autommically be added to the diesel loading following a LOCA. The loading calculation provides a total load for all automaticaliy loadad MOVs at 33 kW which is rounded to the nearest 25 kW or 50 kW. The load associated with the MOV is 0.22kW. Therefore, the addition of this load does not adversely affect the diesel generator. MOV diagnostic testing will be performed after completion of the modification to ensure correct MOV operation. Therefore, the proposed modification is consistent with the electrical and control systems design basis for automatic containment isolation valves and will provide assurance the MOV will perform its safety function when required. FAILURE MODE AND EFFECTS ANALYSIS The proposed modification was reviewed for potential failure modes and the effects of such failures. These failure modes include:
- The breach of the pressure boundary,
- The plugging of the leak-offline,
- The failure of 1E12-F0738 to close, and
- The failure of 1E12-F074B to close.
The leak-off line has been designed in accordance with the applicable ASME codes. 3 Furthermore, the breach of the pressure boundary for pipes less than or equal to 1" nominal diameter is not a postuitted event. Therefore, there is no concern for the postulated effects of such a failure. Regarding plugging the %" vent line, should this occur, the pressure in the line will increase. The existine, oressure transmitter would alarm if pressure reaches the alarm setpoint. If pressure continues to increase, then the SDC header relief valve will open and protect the piping from the overpressure condition (identical to the current configuration).
Attachment 1 PY-CEl/NRR-2362L Page 10 of 15 sThe failure of 1E12-F073B to close was described in the Section, "SDC Operational Impact." The failure of 1E12-F074B to close was described in the Section, " Failure to Close 1E12-F00748." THERMAL-HYDRA'JLIC CONSIDERATIONS in the proposed piping configuration, the leakage path, which originates at reactor pressure and temperature, will be as follows:
- Leakage from the reactor vessel past the 1E12-F009 and 1E12-F550 valves in parallel,
- Leakege proceeds through valve 1E12-F008,
- Leakage proceeds through the 20" SDC header,
- Leakage proceeds up through the 8" RHR flush line,
- Leakage proceeds up through the proposed %" leak-off line (which is the high point of the system) to the 1" RHR vent line penetration, and
- Leakage then proceeds through valves 1E12-F073B and 1E12-F5588 and continues down the 1" inch vent line to the bottom of the suppression pool in the ,
drywell weir. The phenomena that could affect the physical properties of the leakage includes: flashing as it proceeds to lower pressure sections, condensing as it mixes with the cooler water in each region and rejects heat, and further flashing and condensing as it changes elevation. To assess the thermal / hydraulic performance, the configuration was modeled on a transient, thermal / hydraulic computer code that can assess two-phase flow scenarios (RELAPS). The results of the analysis showed high operating temperatures and fluid velocities in the proposed leak-off line at valve leakage rates even when the leakage rate is below the current Technical Specification 3.4.6 limit of 0.50 gpm per nominalinch of valve size. The elevated operating temperature could cause pipe stress problems and the high velocity steam flow to the suppression pool could cause water hammer problems on a sudden closure of 1E12-F0738. Calculations have shown that if the allowable leakage through the reactor loop suction valves is reduced, the aforementioned problem would be avoided. Therefore, to alleviate this issue the leakage rate through Penetration P421 (the reactor recirculation loop suction valves to the RHR system) is being administratively limited in the surveillance test program to s 0.30 gpm total leakage. Since both the inboard and outboard barriers are maintained closed, i.e., no single failure concem, the limit is being established by requiring that at least one of the two barriers have a leakage rate set s 0.30 gpm. The existing Technical Specification acceptance criteria would be applied to the other barrier. This means that either the sum of leakage through 1E12-F009 and 1E12-F550, or the leakage through 1E12-F0008 must be s 0.30 gpm. At this leakage rate, the discharge to the suppression pool will be a liquid and the operating temperature in the system will be on the order of 190 F to 220 F. These temperatures will have no impact upon the RHR system or components within the near vicinity of the leak-off line (refer to the section entitled,
" Evaluation on Adjacent Components", below).
1 4
P
' ~
- _ Attichment 1 l PY-CEl/NRR-2362L Page 11 of 15 t
1 EVALUATION ON THE FIRE PROTECTION PROGRAM l The impact of the proposed modification upon the PNPP Safe Shutdown capability (Appendix R) was evaluated. The impacts include: 1.) the need to ensure that 1E12-F073B i or 1E12-F620 is closed during shutdown cooling operation, and 2.) the potential impact of fire induced spurious valve operations in accordance with Generic Letter (GL) 86-10, " Fire
- Protection."
With the installation of the proposed leak-off line, the path from the reactor coolant system to the suppression pool established via 1E12-F620 and 1E12-F073B presents the need to isolate the path during safe shutdown operation.. Therefore, it is appropriate to ensure that either 1E12-F620 or 1E12-F073B is closed prior to initiating SDC operation under conditions cssociated with safe shutdown operation. Since the closure of 1E12-F073B or 1E12-F620 is required only under conditions of SDC operation, the conditional manual operation of 1E12-F073B or 1E12-F620 is classified as a
" cold shutdown manual action". As such, this operation is not highly time-sensitive, and need not be implemented until prior to placing SDC in operation. In accordance with {
10CFR50, Appendix R, it is necessary to demonstrate the capability to achieve cold shutdown conditions within 72 hours following a postulated fire. Consequently, the specific time at which the SDC Mode of operation is initiated is not specifically defined. It can be expected that SDC would not be placed into service until well after 8 hours following the postulated fire. The operation of 1E12-F0738 or 1E12-F620 is feasible to implement without adverse impact on the post-fire operator workload, based on the classification of these actions as " cold shutdown" related. Ample time is available for closure of 1E12-F073B or 1E12-F620 prior to initiation of SDC operation, hence there is no impact on the safe shutdown timeline. In accordance with 10CFR50, Appendix R, Section Ill.L, emergency lighting to support operation of either 1E12-F073B or 1E12-F620 is not required (any 8-hnur-rated, battery-
- backed emergency lighting units would be exhausted, and would provide no contribution to l the shutdown effort). This is consistent with the established PNPP position regarding emergency lighting for cold shutdown operations. Hand-held lighting units would be l
credited, in addition to any fixed lighting that may be restored to service as part of the post-fire recovery effort. Therefore, the installation of additional battery-backed emergency lighting units to support the " cold shutdown" operation is not warranted. GL 86-10 states that multiple valve failures need not be assumed provided that circuit ; analysis demonstrates that there are no interlocks or commonalties that could result in simultaneous spurious actuation of the valves as the result of a single fire-induced fault. In the existing configuration, the RHR vent path though valves 1E12-F0738 and
' 1E12-F074B cannot be opened by a fire-induced spurious actuation. This is based on the l fact that the valves are normally closed and that there are no interlocks or commonalties I
between the two valves. Hence, there would be no impact upon the initiation of RHR-B SDC if the scenario required its use. However, since the proposed configuration has 1E12-F073B being normally open, a vulnerability due to a single fire-induced spurious actuation might have been created. An evaluation was performed which resulted in this not being a potentialissue. For any scenario for which the use of RHR-B SDC is credited, the control and power cables t
Attichment 1 PY-CEl/NRR-2362L Page 12 of 15
. associated with RHR-B SDC equipment will be free of fire damage. For any scenario for which the 1E12-F073B or 1E12-F074B cables are exposed to potential fire damage, the use of RHR-A SDC is credited. However, if the plant is being shutdown with RHR-A, the 1E12-F620 valve may need to be manually closed should a fire-induced spurious actuation cause damage to 1E12-F073B. In case of a fire in the RHR "B" heat exchanger room, the 1E12-F620 valve could be manually closed, if necessary, after the fire has been mitigated. If the room is totally inaccessible, the attemate shutdown cooling flow path can be utilized (this flow path does not use the common RHR piping that contains the leak-off line). Therefore, plant cold shutdown in accordance with 10CFR50, Appendix R can still be achieved. The proposed modification has no impact on combustible loading, or any active or passive fire protection feature.
In conclusion, the changes to the Fire Protection Program created by the proposed modification do not adversely affect the ability to achieve and maintain safe shutdown. OTHER CONSIDERATIONS INADVERTENT REACTOR VESSEL DRAIN DCWN The evaluation of the proposed modification considered the possibility of draining the vessel during shutdown conditions. The inadvertent opening of 1E12-F073B during an outage could allow reactor coolant to drain to the suppression pool. In accordance with the PNPP i Program for Shutdown Safety (PAP-0116, " Shutdown Safety") this configuration is not considered a new Operation with a Potential to Drain the Reactor Vessel (OPDRV). The inner diameter of the new branch is s 1.53 inches and the 1E12-F073B valve will automatically close when the reactor vessel reaches Level 3. EVALUATION AGAINST INTERNALLY GENERATED MISSILES As stated earlier, the design basis for the SDC header includes compliance with GDC 4 as it relates to protecting the system against internally generated missiles. The proposed isak-off line was reviewed to ensure that it is not subject to damage from an internally generated missile. For high-pressure piping, USAR Section 3.5.1.1 states that credible missiles are limited to temperature or other detectors installed on piping or in wells that could become a missile if a single circumferential weld failure would cause their ejection. Since there are no missiles identified for the RHR "B" room, the proposed leak-off pipe will not be subject to damage from intemally generated missiles. EVALUATION FOR PROTECTION AGAINST PIPE BREAKS (JET IMPINGEMENT AND PIPE WHIP) As stated earlier, the design basis for the Shutdown Cooling header includes compliance with GDC 4 as it relate.s to being protected from the environmental and dynamic effects of postulated accidents (jet impingement and pipe whip). The leak-off line is in the RHR "B" heet exchanger room which is an area that was previously evaluated for high-energy pipe breaks. The review of this room showed that the Reactor Core Isolation Cooling Steam Supply Line is the only line in the area with postulated breaks. Based on the review, it was I concluded that the proposed leak-off line and the extended SDC header boundary is not routed in a manner that would be a target for jet loads or pipe whip from these postulated pipe breaks.
]
i Attachment 1 PY-CEl/NRR-2362L Page 'i3 of 15 s EVAlt,JATION ON ADJACENT COMPONENTS As stated in USAR Sections 3.6.2.1.1 and 3.6.2.1.2, fluid systems are defined as either Moderate Energy or High Energy fluid systems. The SDC header is considered a Moderate Energy system since it does not operate at a condition either above 200 F or at a pressure over 275 psig for more than 2% of the operating life. Considering the proposed leak-off line and its operation, the operating temperature of the system may be slightly in excess of
' 200*F when the leakage is at the proposed administrative limit (the supporting analysis shows a temperature ofless than 220 F at 0.30 gpm leakage). With this change in operating temperature, the SDC header could now be defined as a High Energy fluid system. However, in this operating mode, the SDC subsystem piping is isolated from the high energy fluid contained in the reactor vessel during normal plant operation via closed valves 1E12-F008,1E12-F009, and 1E12-F550. USAR Section 3.6.2.1.3, states that piping systems that fall into this category are exempt from the consideration of postulated pipe breaks. Therefore, the change to the operating conditions as a result of the leak-offline addition does not introduce any new pipe break concerns.
The postulated pipe break scenario for both High Energy and Moderate Energy fluid systems (USAR Section 3.6.2.1.6) exempts consideration of breaks or cracks for pipes less than or equal to 1-inch nominal diameter. Since the new leak-off line is %" piping, it is not required to postulate either a crack or a pipe break or to evaluate any dynamic effects or environmental effects on the adjacent safety-related equipment in the RHR "B" heat exchanger room. From an environmental perspective, the proposed leak-off line will be uninsulated and will reject a small quantity of heat to the RHR "B" heat exchanger room. The environmental conditions have been evaluated and found to have negligible effects upon the environmental qual;fication of the safety-related equipment located within the room. EVALUATION ON THE GL 96-05 PROGRAM NRC Generic Letter 96-05, " Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves", addresses potential safety concerns involving MOVs. The
. PNPP GL 96-05 MOV Program addresses the recommendations of the NRC Generic Letter.
Valve 1E12-F073B was previously exempt from this program as it was maintained normally closed, was not required to change position for any design basis event, and received no automatic isolation signal. Implementing the proposed modification,1E12-F073B will become a normally open valve with an automatic isolation signal. Hence, it will be includer; in the PNPP GL 96-05 MOV Program. Analyses were performed which demonstrate the valve would be capable to close against the dynamic loads associated with the system leakage. EVALUATION ON THE SUPPRESSION POOL The proposed leak-off line which routes leakage from the RHR Shutdown Cooling header to the suppression pool will not have an adverse effect on the suppression pool level and temperature. In the current plant configuration, leakage past 1E12-F008,1E12-F009, and 1E12 F550, which is allowed to be as high as 5 gpm, is routed to the suppression pool via the SDC header over-pressure relief valve,1E12-F005. Afterinstallation of the proposed leak-off line, the leakage will still end up in the suppression pool, it is simply routed via a i j
r; Anachm:nt i i PY-CEl/NRR-2362L Page 14 of 15 l l l 1different flow path. Since the current leakage terminates in the suppression pool and level is i not adversely affected, a change in the leakage path will not affect the suppression pool. In ' addition, since the proposed modification administratively limits the allowable leakage flow below that which is currently permitted, the heat input to the suppression pool will be ! reduced. CONCLUSION i The proposed modification installs a leak-off line in the SDC portion of the RHR system to l control pressure caused by valve leakage. The modification has been designed, and will be manufactured and !nstalled in accordance with the original RHR system codes and l standards. The modification will not impact any safety function of the RHR system. The ; reliability of the RHR system has been improved, with the modification eliminating an operator work arcund which will significantly reduce the collective dose to plant operators. Therefore, installation of this modification is appropriate. ENVIRONMENTAL CONSIDERATION l l The proposed Technical Specification change request was evaluated against the criteria of 10 CFR 51.22 for environmental considerations. The proposed change does not significantly increase individual or cumulative occupational radiation exposures, does not ! significantly change the types or significantly increase the amounts of effluents that may be released off-site and, as discussed in Attachment 2, does not involve a significant hazards consideration. Based on the foregoing, it has been concluded that the proposed Technical l Specification change meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement. l [ l l t
Attachment i PY-CEl/NRR-2363L Page 15 of 15 Figure 1 Diagram of RHR Leak-Off Line Modification Contairanent Drywell e n' I J D0007B
\ h RHR Vent /[
E12-F074B
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/ I N'1 . I I '
E12-F073B E12-F558B l ! RHR HX, IE12B001B >VVVL nf E12-F620 Suppression Pool New %" leak-offIJne from SDC Header to Suppression Pool
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O Condensate _)i(__q ' Transfer ' E12-F315 E12-F020 Reliefto Suppression Pool 4 E12-F005 p' p-1 SDC Suction E12-F005A ( ' ' )k,' from Rx Recire E12-F008 E12-F009 Line 7 g
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E12-F550 Suction Lines 7-~ Existing 8" fill / vent on I bottom of 20 inch SDC _ _d [ \E12-F006B Header to remain i
Attichment 2 PY-CEUNRR 2362L Page 1 of 2 SIGNIFICANT HAZARDS CONSIDERATION t The standards used to arrive at a determination that a request for amendment involves no significant hazards considerations are included in the Nuclear Regulatory Commission's Regulation,10 CFR 50.92, which states that the operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident praviously evaluated; or (2) create the possibility of a new or different kind of accident from any previously evaluated; or (3)
- involve a significant reduction in a margin of safety.
The proposed amendment is requesting Nuclear Regulatory Commission review and approval of changes to the Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report (USAR) to incorporate descriptions (in the form of text, tables, and drawings) of a
. modification to the plant involving the installation of a leak-off line in the Residual Heat Removal (RHR) System. The leak-off line with its attendant supporting valve changes is designed to increase the overall reliability of the RHR system and to eliminate undesired manual operator actions.
The proposed amendment has been reviewed with respect to these three factors and it has been determined that the proposed change does not involve a significant hazard because:
- 1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed modification has been designed, and will be procured and installed in : accordance with the original design codes and standards. The safety functions of ) the RHR system have not been. impacted by the change. Systems supporting the ! operation of the RHR system have not been affected by this modification. Though , the modification affects the Containment System, the containment remains capable j of performing its associated safety functions to the same level as the original design. ' The accidents of concern are the Loss-Of-Coolant-Accident (LOCA) and the Loss of Shutdown Cooling. The proposed change has been designed in accordance with the original codes and standards. The proposed change will not alter the operation l of any plant equipment assumed to function in response to the aforementioned analyzed events or otherwise increase their failure probability. Therefore, the probability of occurrence or the consequences of an accident previously evaluated remains unchanged.
?.. The proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.
The proposed modification has been designed, and will be procured and installed in accordance with the original RHR system design codes and standards. RHR system functions have not been impacted by the change. Systems supporting the operation of the RHR system have not been affected. Failure of the modification to perform its design function due to leak-off line failure or blockage would be identical to the
Attachm:nt 2 PY-CEl/NRR-2362L Page 2 of 2 current RHR system performance, improper operation of the valves associated with the modification have been eva;uated and will not prevent or otherwise inhibit the RHR or Containment systems from performing their applicable safety functions. Missile generation is not a concern since no mechanisms conducive to missile generation have been introduced. Electrical analyses have shown there is no adverse effect upon the diesel generator loadings. A single failure of the new configuration will not result in more than the loss of a single RHR loop which is already analyzed. Therefore, the possibility of a new or different kind of accident from any previously evaluated has not been created. 3.' The proposed change will not involve a significant reduction in the margin of safety. The proposed modification has been designed, and will be procured and installed in accordance with the original RHR system design codes and standards. The RHR and Containment systems remain capable of performing their safety functions. Systems supporting the operation of the RHR system have not been affected. Hence, the RHR system margin of safety with respect to safety classification, l protection, redundancy, anc seismic classification remains unaffected. I I The margins of safety contained in the Technical Specifications and the associated l Bases also remain unaffected by this modification. Specifically, Technical l Specifications 3.4.6, " Reactor Coolant System Pressure isolation Valve Leakage"; 3.4.9, "RHR Shutdown Cooling System - Hot Shutdown"; 3.4.10 "RHR Shutdown Cooling System - Cold Shutdown"; 3.6.2.1, Suppression P+ol Average Temperature"; and 3.6.2.2, " Suppression Pool Water Level"; and the associated Bases remain unchanged and fully applicable. Hence, the margins of safety defined in the Technical Specifications remains unaffected. Therefore, the proposed modification does not involve a significant reduction in the margin of safety. l l l
Attachment 3
) PY-CEt/NRR-2362L Page 1 of 12 The RHR system is connected to higher pressure piping at shutdown
' shetion, shutdown return, LPCI injection, head spray, and heat exchanger
~
supply. In general, pressure interlocks prevent opening valves to the lov pressure suction piping where the reactor pressure is above the shutdown range. These same interlocks initiate valve closure on increasing reactor pressure. In addition, a high pressure check valve vill close to prevent reverse flow if the pressure should increase. Relief valves in the discharge piping are sized to account for leakage past the check valve. Shutdown suction has two gate valves, F008 and F009, in series which have independent pressure interlocks to prevent ( opening at higher inboard pressure for each valve. No single active failure or operator error vill result in overpressurization of the low pressure piping. In vent of leakage past F008 and F009,
.PT-N057 provides indication and alarm to the control' room operatorg
~
if theleak-ollineisisolatedanddin AddM a t The shutdown return line T a T ing check valve, F050, to protect it i
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from higher vessel pressures. Additionally, a globe valve, F053, is located in series and has pressure interlock to prevent opening at high i'nboard pressures. No single active failure or operator error vill cause overpressurization of the lover pressure piping. The LPCI injection line has a piston check valve, F041, and a high-low pressure interlock,'to protect it from higher vessel pressure. No single active failure or operator error vill cause overpressurization of the lovei pressure piping. The head spray line has a sving check valve, F019, to protect it from higher vessel pressure. Additionally, a-gate 6 valve, F023,.is located in series and has pressure interlocks to prevent opening at higher inboa'rd pressure. No single active failure or operator error vill cause overpreIssuriration of the lover pressure piping. ' The heat exchanger. steam supply line has 'a globe valve for shutoff, F052. The operator admits' steam through F052 and sets the pressure regulating valve, F051, to limit heat exchanger pressure. Also,'F087 THIS C.@ 5.4-46 line to the suppression pool
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ecoos tJ v N Po ' > e (Rev. 7 3/95) O PERRY HUCLEAR POWER PLANT THE CLEVELAND ELECTRIC ILLUMINATING COMPANY Residual Heat Removal System Figure 5.4-13 (Sheet 2 of 3) (Dvg. D-302-642)
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Attachment 3 PV-CEl/NRR-2362L l Page 4 of 12 l
- 3. Residual Heat Removal Heat Exchanger Vent and Relief Valve Discharge Lines (P118/P133, P107/P109, P431/P421, and -
P429/P419) I Residual heat removal heat exchanger vent lines discharge to the suppression chamber. Two normally closed, remote-manually actuated, raotor operated valves outside containment, and a check valve, located between rnntainment and the dryvell, provide isolatio except for P431. THIS CR. d
& Relief valve discharge lines from the residual heat removal 4
4 heat exchangers and various emergency core cooling system Y suction and discharge lines discharge to the suppression pool. These vent lines are isolated by the relief valves. The addition of block valves vould defeat the purpose of the ! relief valves. The relief valves set pressure is greater than 1.5 times containment design pressure. l i i A postaccident sampling system sample return line also ties into the relief valve discharge header and is isolated by two locked closed, remote-manually actuated solenoid valves outside containment. (
- b. Effluent Lines from the Suppression Pool (P103/P103 and P401/P401 an'd P101/P101; P102/P102 and P402/P402 dnd P403/P403)
The low pressure core spray, high pressure core spray, reactor co,re isolatir. cooling, and residual heat removal suction lines are equipped with remote-manually actuated, motor operated gate valves outside containment. These valves provide the ability to isolate in the event of a line break and also provide long term leakage , control. , 7 - m Penetration P431 has one normally closed, remote-manually %ated, motor operated valve and one normally open, automatic closing, remote-manually actuated, motor operated vain o.ztside containment, and a check valve, located between containment an'd the drywell. 6.2-100 ' T Hi S C. R
Attichment 3 fg PY-CEl/NRR-2362L Page 5 of 12 T;&E 6.2-32 (c:ntimed)
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G G G AI mg Pwr. Noca. Tirnetsee) (9) " T10) 1E his Dir. o - m m m sed. 2 m Revision 8 m 6.2-208 Oct. 1996
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PY-CEl/NRR-2362L Page 6 of 12 i TABLE 6.2-36 PRIMARY CONTAINMENT ISOLATION FOR LINES THAT PENkrxATE CONTAINMENT AND CONNECT TO THE SUPPRESSION POOL Isolation Device (s) III Line ' Outside Reference Containment _Section Influent Lines Low Pressure Core Spray Minimum Flow Line MOV 6.2.4.2.2.2.a.1 Low Pressure Core Spray Test Line MOV 6.2.4.2.2.2.a.1 ! High Pressure Core ' Spray Minimum l Flow Line HOV 6.2.4.2.2.2.a.1 High Pressure Core Spray Test Line MOV 6.2.4.2.2.2.a.1 ( Residual Heat Removal Suppression Pool Cooling and Test Line MOV 6.2.4.2.2.2.a.1 Residual Heat Removal Steam Condensing Mode Bypass Line MOV 6.2.4.2.2.2.a.1 Reaidual Heat Removal Minimum Flow Line MOV 6.2.4.2.2.2.a.1 l Reactor Core Isolation Cooling Minimum Flow Line MOV, CV 6.2.4.2.2.2.a.2 i Reactor Core Isolation Cooling Turbine Exhaust Line MOV, CV, RV 6.2.4.2.2.2.c.3 Residual Heat Removal Heat Exchanger Vent to Suppression 74,[l$ CR Pool MOV, 6.2.4.2.2.2.a.3 Relief Valve Discharge . Lines RV, MOV 6.2.4.2.2.2.a.3 l
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l Attachment 3 o -' PY-CEl/NRR-2362L l Page 7 of 12
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PERRY HUCLEAR POWER PLANT FOR USAR USE ONLY NUCLEAR SAFETY RELATED M WINA m cC NPANY Containment and Drywell Isolation Figure 6.2-60 (Sheet 3 of 4) (Dwg. D-500-763) i y b
an 1 Attachment 3 PY-CFl/NRR-2362L J Page 8 of 12 "aa a .a Pass fins 8
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yy . NOTES-89 ## t.7 8. !
. ig! SEE REFERENCE S FOR SUPPLEMENTAL FLOWS ENTERmG h
DOWNSTREAM OF EI2-F050 DURINO NOlth4AL PLANT OPERATIONS. *.
- 2.
PIPING 8ETWEEN POmTS WITH EMPTY DATA BLANKS (SEE ALSO TABLE 3) SHALL BE SIZED BY CUSTOhER OR AE BASED ON
- 19. FLOW.SHOWN is A MAJGMUhL ACTUAL PLOW WILL BEINDICATED 1.ATER FOR EACH PROJECT. f 3
SPECIFED OPERATING CONDITIONS. EMPTY DATA BLANKS CAN DE 20. ' MAJGWUM SHELL SIDE PLOWILATE 18 7800 OPhL 3 g FILLED IN BASED ON ACTUAL ARRANGEMENT OR EQUtVALENT 21. FLOW SHOWN AT POSITION ft DOES NOT INCLUDE FLOW FROM FUEL 0
- HYDRAUUC DATA SUBhuYFED TO BWASD POR REYlEW. POOLCOOUNG ANDCLEANUPSYSTEnt b
@ ~ INDICATES THE DATA IS NOT 840NIFICANT. 22. SEE SYSTEM DATA SLEET POR SUOOESTED YALVE SIZING. O O 3. 21 7 SHOWN AS TYPICAL FOR ONE LOOP. IF 1DOPS Oh SIDE 1 AND SIDE It SUCTION TEMPERATURE MD PPM ARE FOR LOOPS A&S m' ARE NOT SYMME11UCALLY ARRANGED. VAT.UES FOR DOTH SIDES LOOP C CONDITIONS ARE O PSIG YESSEL PRESSURE 125'F.
A SHALL BE SUBhuTTER 24 THE HX INLET PRES $URE SHALL BE OREATER THAN 40 PSIA TO l 4 M 4. t AH VALUES FOR EQUIPMENT WITHIN GE SCvPW ARE AS NOTED.
- S.
ELEVATlONS ARE NOT IN"LUl**.3 m AP VALUES OlVEN. ELEVATIONS 21 knNthEEE THE POSSSE.!TY OF PLOW.fMDU FOR LOOPS A AND B, MODE O MAYBE ILDIENATED FROM cp' ' SHALL SE INCLUDED WHEN DETERMINING FINAL VALUES FOR THE CONSIDERATION DuluMO SHUTDOWN COOUNG F MOV-POM 85 ' EMPTY DATA BLANKS. ELECTIUCA11Y DISABt.ED
' 6. .
= INDICATES MAXIMUM (X) AND MINIMUM (Y)'VALLT.S F MOOE SPECIm WHEN SHUTDOWN COOLING 18 INITIATED IN THE A*S LOOP ONLY
- 7.
DASHED LINES INDICATE FLOW DOES NOT PASS THRU THESE POINTS ONE VALVE SHOULD SE DLSASLED AT ANY OIVEN TibE CUSTOMER 01
. SOLID LINES INDICATE FLOW DOES PASS THRU THESE POtNTS.
ESTABLISHED DESIGN ALTERNATE TO GE STANDARD. :l {' S. THE NPSH AVAILASLE IN MODES A-2. B.t. & D-l. AT A REFERENCE 7 . LOCATION 2 FEET ASOYE THE PUMP MOUNTING FLANGE MUST . 26. THE NEW STRAINER HAS BEEN DESIONED TO ENSURE ADEQUATE A EQUAL OR EXCEED $ FT. ASSUMING SATURATION TEMPERATURES OF PUMP NPSH UNDER MAXIMUM POSTLR.ATED LOADDIO RESULTWO 2124 AND 3S8T RESPECTIVELY. THE NPSH AVAILABLE AT THE PUMP FROM LOCA<sENERATED AND PRE.LOCA DEBRIS MATERIALS AND A """ SUCTION NOZZLE MUST SQUAL OR EXCEED TMis VALUE PLUS THE FOR A MAXIMUM EXPECTED SUPPRESSION POOL TEMPERATURE OF - DIFFERENCE IN ELEVAT10N BETWEEN TlE REFEstENCE LOCATION 1854. THE REQUIREMENTS FOR ONE STRADER 30% PLUOGED (SEE "" AND THE CENTER-LINE OF THE PUMP SUCTION NOZZLE. MODE DESCRIPTIONS) AND NPSH REQUIREhENTS AT 2124 (SEE NOTE -
. 9. PIPING SYSTEM DESION PRESSURE AND TEheERATUltf AND THE 8)NO LONGER APPLY.
ESTIMATED LINE SIZR8 ARE POR INFORMATION ONLY, ACTUAL DES ON PRESSURE AND TEMrERATURE AND UNE StEES AS .17. NNN gg , g DETEmmED sY elPDeo DESiONER SHALL hEET THE PROCESS DATA
's. HYDRAUuc REQUIREMENTS. REFER TO HARDWARE DWOS. FOR - Qd j Wa f
.7 M /S I:; pit-NOZZtE SIZES ON RESuPruBD mQUiPhENT. -
7 8 10. FUEL POOL CONNECTIONS MUST PROYlDE ADEQUATE NPSH TO ofSRpry gd. 9%gA6Aaretst$ g>J o A . AVOlD PUMP CAYITATION AND ATTHE SAh5 TIME PROVfDE FOR GREATER THAN Mm1 MUM PUMP P14W. "TM $HQTE18Wb$ QBW P"O M M Ni - a L TAst.E I DGICATES VALVE POSmON DURmo VARIOUS MODES OF g,gggpp (,,lQg plpMdn QURlba6
. OPERATION.
- 12. DELETED.
N0blEMbb S'i
- 11 TYPICAL VALUES POr, max. SUPPRESSION POOL TEMP SHOWN. * * ,
FMAL TEMPERATURE DEPENDS ON INITIAL POOL WATER ' TEMPERATvRE AND POOLWATER v0LUMr.
,; , PERRY NUCLEAR POWER PLANT H. WATERFLOWS AREINGPM STEAMPIAWS AREIN1000LB14st THE CLEVELAND ELACTRIC IS. MAXtMUM SOH Teo FEET.
- ILLUMINATING COMPANY j
- 14. SERVICE WATER CROSSTIE SHALL SE SIZED TO FLOW 200 @M AND
? ENOUGH HEAD TO PLOOD THE CONFAl>4EENT {
- 17. TIE WEIGHT OF WATERINTHE $6RffDOWNCOOUNG SUBSYSTEM f s
Pirms, mCouDmo THE HEAT EXCHANOERS AND PUMPS SHAU. NOT . Res.i dual Heat Removal System
- EXCEED 2n.coe Las AT TPF TO PREVENT DlumON OF SteD8Y uQuu CONrROL NEUTRON AnSORsERsELOWMmlMuM process Diagram i
~ REQUUuiMENTS i' Figure 6.3-3 (Sheet 1 of 3) J F e
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Attachm:nt 3 PY-CEl/NRR-2362L Page 9 of 12 THIG Cff-The upper pool shutdown cooling valves E12F037 A, , the two series RHR heat exchanger vent valves E12F073 A nd F074 A, B and the RHR shutdown cooling mode suction valves E12F006A, B are all normally closed and thus require no automatic close sional for system g a { 1E12F0738 is normally open%(RHR heat exchanger vent valve' and thus requires an automatic signal to close. The LPCI pump motors and injection valves are provided with, manualoverridec5ntrols. These controls permit the operator 3 to manually control the system subsequent to automatic initiation. 7.3.1.1.2 THtS CR Containment and Reactor Vessel Isolation Control System (CRVICS) - Instrumentation and Controls
- a. CRVICS Function k g
I The CRVICS, also known as nuclear steam supply shutoff system (NSSSS), includes the instrument channels, trip logics and actuation circuits that automatically initiate valve closure providing isolation of the containment and/or reactor vessel, and initiation of systems provided to limit the release of radioactive materials. l 1 j See Section 6.2.4 and Table 6.2-32 for a complete description of primary containment and reactor vessel process lines and isolation , signals applied to each. i The Technical Specifications require that \ l several CRVICS Instrumentation channels for the Main Steam Line Isolation Valves meet response time criteria. Table 7.3-1 provides j the acceptable response for these channels along with any clarifying information. j 7.3-15 Revision 8 Oct. 1996
Attachment 3 PY-CEl/NRR-2362L Page 10 of 12 s ; magn. . 1 teen TEST 50 led 0lO PILOT 15 ENERG12EB. THE M6tN STEAM 150LATions VALVE OPERATipt 15 SLimit Y EXHAUSTED (60 SEC CLD5 Mile TWE ) iMILE VALVE IS CLOSED BY ACTlose GF THE VALVE SPRING WlTHOUT &ID OF AtR IW55481E. 2. THE ALARMS AND VALVE INotCATING Lle4TS SHEksi 138 TM FCD ARE SYSTE" REQUIREMEN IN ADDITitel TO THOSE SHlhei ON TM STSTEM P<. A00lTicasAL INFORMAT 10N ON AL ARMS, VALVE P05 til0N IN01CATING LIGHTS. Aas PROCTSS INSTRUMENT &Tl004 NOT SHOuM ON THIS FCD MAY BE OSTAINED FROM .msspL Coc aaasstimav as&Avs& Oswicas Ana Mer seemns em Ptp 33t.gpr usugas maeuineD1D CA.AmspV MagreeM. X G. ALL EQulPPENT NOTED 4e80 INSTRtMENTS ARE NETIEED BY SYSTEM NO. 821. unsLE55 OTHERWISE
- 4. 1Ms(AsaJuauptsaclab pusep =M-AE3tMESRfE.15 LISED IN AD$ LOGIC A AE ONLY.
G. 'EACM SAFETY / RELIEF VALVE 6 AUTO DEMIC550RIZAT104 CONTRG. LOGIC CIRCUlf SMA HavE AUTCREDUNDANT DEPRESSIRIZAT POWER10N FUMSUP Tilpi. IES S0 TH&I A SINGLE FAILURE WILL NOT
- lSA8LE T do. THE NUCLEAR BOILER SYS SHALL 8E DESIGIED IN ACCORDANCE WITH "P10 POSED CRITERIA APy,1CASLE.To FOR NUCLEAR THE CONTRG. PedEA PLAssi PROTECTION SYSTEM IEEE 279" AS CIRCu!TRY. -- - - . - - - - - - -
7 ISOL! TION LOGIC SHALL SE "TAILSAfT".1.E., LOGIC SHALL BE DESIGNEC T0 INITI ATE ISOLA?10N FUsicTIONS t*EN DE-ENERGIZEB.
- 8. DELET ED 9.
T>E DEVICES ESSENTIAL AND MUSTIN MEET THIS ARE) tHE REQUIREfEIITS OF IEEE 279.A5 WELL A5 OTM
- 10. DELE.T E.D IL ALARM FRIM ROTARY CONTACT. -
#2. DELETE.D
- 11. THE SRf5 RECEIVE OVEWPRE5'5URE TWlPS FWOM VARiopS TRIP UNITS. FOR THE TRFP UNIT ASSIGNMENT FOR EACH SRY 5 EE TABLE 5 OF SUPP'LEMENTAL. DOCUMENT' 1. .
K SRV MONfTOR SHALL CAUSE A90NUIC8ATION IF AftY SRV LINE DISCHARGE PRESSURE EXCEEDS SETPoestT. k SEE Rap 11 POR REA40TE $NUTOswsJ REeutREMEMTS.
- 16. SEL. SIhrITCH IS 2 POSITION
- WORW"
- TEST *, WAINTAINED CONTACTS ,
HEY REMOVASLE fN 'MORM" P9ErT"10N. . v- {
- 17. TYPICALLY DIVISION 1 ISOLATION LOGIC CONTROLS THE OUTBOARD VALVES AND DIVISION 2 ISOLATION LOGIC COPffROLS IIIE P40ARD VALVES. HOWEVER THE OPPOSITE IS PERMISSIBLE PROVIDED THE CONDITE.O OF DIVISIONAL SEPARATION ARE NOT VIOLATED.
=__ _ _ _ - _
(Rev. 6 3/94) e PERRY NUCLEAR POWER PLANT THE CLEVELAND ELECTRIC ILLUMINATING COMPANY Nuclear Boiler System Figure 7.3-3 (Sheet 1 of 7) [Dwg. D-808-303(1)]
Attachm:nt 3 PY-CEl/NRR-2362L Page 11 of 12 7 - - - - -- - - - - --. - - - _ 3 i I w ca~ ~ I i lr, .
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m % roa mamano cmmoe : .:st: wrm oevice son, anasa,)) NUCLEAR SAFETY RELATED (Rev. 5 3/93) O PERRY NUCLEAR POWER PLANT TNE CLEVELAND ELECTRIC ILLUMINATING COMPANY Nuclear Boiler System Figure 7.3-3 (Sheet 2 of 7) IDwg. D-808-303(2)]
Attachmsnt 3 PY-CEl/NRR-2362L Page 12 of 12
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ggt tacht t 1 (Rev. 7 3/95) 1 PERRY NUCLEAR POWER PLANT THE CLEVELAND ELECTRIC ILLUMINATie4G COMPANY Residual Heat Removal System Figure 7.3-5 (Sheet 4 of 5) [Dvg. D-808-309(4)]}}