ML18026A463
| ML18026A463 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/20/1996 |
| From: | Byram R PENNSYLVANIA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML17158B557 | List: |
| References | |
| PLA-4428, NUDOCS 9603250106 | |
| Download: ML18026A463 (36) | |
Text
CATEGORY 1 REGULATE.i.: I NFORMATIGA DI'O'PRIE JTIOh~VSTEN (R I DS ) '- .'+i'; ~
f ACCESSION NS(3:9603250106 'OC:DATE: 96/03'/20 NOTARIZED: YES .DOCKET If FACIL:50-381 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 S <;quehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION BYRAM,R.G.
RECIP.NAME Pennsy'vania Power S Tight Co.
RECIPIENT AFFILIATION 5~
Document Control Branch (Document Control Desk +~+p i~ 7e 4 5'1'~
SUBJECT:
For"yards application for amends to licenses NPF-14 & NPF-22, revising TSs re HPCI pump automatic transfer to suppression poo! logic e imina t i on.
~ A TITLE: OR
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DISTRIBDTION CODE: AOOID COPIES RECEIVED:LTR Su':.a.ittal: General Distribution I ENCL j SIZE: T E
NOTES: 05000387 RE('1'PIENT COPIES RECIPIENT COPIES I'> .')DE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD] ~ LA 1 1 PDl-2 PD 1 1 POS! l, NV,C 1 1 INTERNAL: ACR.'~ ILE CENTER 1 1 1 NRR/ 'E/EMCB NRR DRC 1 1 NRR,SSA/SPLB NRR/DSSA/SRXB 1 NUD<- S-ABSTRACT OGC/HDS2 1 0 EXTERNA I 'OA(. NRC PDR 1 1 D
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NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTEl CONTACT THE DOCUMENT CONTROL DESKS ROOM OWFN 5D-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION I ISTS FOR DOCUMENTS YOU DON'T NEEDl TOTAL NUMP OF COPIE.,'D REQUIREt'NC' i8
Pennsylvania Power & Light Company Two North Ninth Street ~Allentown, PA 16101-1179 ~ 610/774-5151 Robert G. Byram Senior Vlcc President-Nuclear 610/774-7502 Fax: 610/774-5019 MAR 20 1996 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Mail Station P 1-137 Washington, D. C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION PROPOSED AMENDMENTNO. 197 TO LICENSE NPF-14 AND PROPOSED AMENDMENTNO. 155 TO LICENSE NPF-22: HPCI PUMP AUTOMATICTRANSFER TO SUPPRESSION POOL LOGIC ELIMINATION Docket Nos. 50-387/NPF-14 and 50-388/NPF-22
References:
I Letter, PLd-3696, H. K Keiser to C. L. Miller, "Submittal of the IPE Report", dated December 13, 1991.
- 2. Letter, Phd-3902, H. K Keiser to C. L. Miller, "Individual Plant Examination Supplemental Inforntation ", dated January 11, 1993.
- 3. Letter, PL4-4196, R. G. Byram to C L. Miller, "Manual Control of HPCI Suction Transfer", dated October 11, 1994.
This letter proposes changes to the Susquehanna Steam Electric Station Units 1 and 2 Technical Specifications that change the OPEN logic for HPCI suction valve HV-155/255-F042 in order to eliminate the HPCI pump auto-transfer on high suppression pool level. Implementation of the proposed change and the associated plant modifications completes installation of modifications identified as vulnerabilities by the Susquehanna SES IPE.
Enclosure A to this letter is the "Safety Assessment" supporting this change. Enclosure B to this letter is the "No Significant Hazards Considerations" evaluation performed in accordance with the criteria of 10 CFR 50.92. The proposed changes have been reviewed by the Susquehanna SES Plant Operations Review Committee and the Susquehanna Review Committee.
Enclosure C to this letter is the current pages of the Susquehanna SES Units 1 and 2 Technical Specifications marked to show the proposed changes.
9603250106 960320 PDR ADQCK 05000387I P PDR
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FILES R41-1/A17-2/A17-17/A-30 PLA-4428 Document Control Desk PP&L plans to implement the proposed changes during the upcoming Unit 2 Refueling and Inspection Outage scheduled to begin in May of 1997 and the Unit 1 10th Refueling and Inspection Outage currently scheduled to start in March of.'1998. Therefore, we request NRC:
complete the review of this change request by March 1, 1997 to support our scheduled implementation dates.
Any questions regarding this request should be directed to Mr. William W. Williams at (610) 774-7742.
Very truly yours, At a ents c'py: NRC Region I Mr. C. Poslusny, Jr., NRC Sr. Project Manager - OWFN Ms. M. Banerjee, NRC Sr. Resident Inspector - SSES Mr. W. P. Dornsife, Pa. DEP
ENCLOSURE A TO PLA-4423 Page 1 of 1 1 SAFETY ASSESSMENT HPCI PUMP SUCTION AUTOMATICTRANSFER TO SUPPRESSION POOL LOGIC ELIMINATION(UNITS 1 AND 2)
The purpose of this proposed change to the Susquehanna SES Technical Specifications is to modify the OPEN logic for HPCI suction valve HV-155/255-F042 to eliminate the HPCI pump auto-transfer on high suppression pool level. The modifications necessary to support this change consist of the following:
Currently F042 receives an OPEN signal for high suppression pool level when the HPCI injection valve is not fully closed or on CST low level. This modification removes the contacts for high suppression pool level and the HPCI injection valve position from the OPEN logic so that F042 can only be opened by a low CST level signal or remotely by hand switch.
The OPEN logic for F042 was previously modified because auto-transfer to suppression pool limited the use of the HPCI system to control pressure following MSIV closure events that do not require HPCI injection into the RPV. That modification added an auxiliary relay in series with the high suppression pool level contact in the OPEN logic for F042 such that the revised logic gives an OPEN signal on high suppression pool level only if HPCI is injecting to the RPV. That relay is removed by this change.
This change increases HPCI system reliability in an ATWS event which is characterized by rapidly increasing suppression pool temperature. Currently in an ATWS event, high suppression pool temperatures'necessitate the manual bypass of the HPCI suction transfer logic allowing the operator to re-align suction to the cooler CST water. The HPCI system is designed for continuous operation with suppression pool water temperatures up to 140 'F. In an ATWS event with no additional failures, suppression pool temperature is expected to reach 179 'F which exceeds the maximum temperature considered in the design of the HPCI system.> For ATWS events which involve additional equipment failures, much higher pool temperatures are expected.
In particular, the calculated MSIV-closure ATWS with SLCS failure results in a peak pool temperature of 274 'F which is well beyond the HPCI design temperature.
The Susquehanna EOP for ATWS mitigation instructs the operator to maintain HPCI suction on the CST and to bypass the high suppression pool suction transfer logic if necessary. Currently, the operator must manually bypass the transfer logic circuitry; that action must be accomplished outside the "control room. For the ATWS with SK,CS failure, the bypass cannot be carried out in time to assure continued operation of HPCI. Should HPCI fail, rapid depressurization of the "Evaluation of Susquehanna ATWS Performance for Power Uprate Conditions", GENE-637-024-0893, p. 9, September 1993.
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ENCLOSURE A TO PLA-4428 Page 2 of 11 reactor is required in order to obtain vessel makeup from low pressure sources. Operation of a critical reactor at low pressure is highly undesirable.
In non-ATWS events, except for Station Blackout, the Susquehanna EOPs permit the operator to manually bypass the suction transfer logic and maintain HPCI suction on the CST only if suppression pool temperature exceeds the HPCI design limit of 140 F,. This suppression pool temperature restriction, which is not part of the generic BWROG Emergency Procedure Guidelines, was added to the EOPs to prevent conflict with the current plant design basis. As discussed later in this Safety Assessment, this temperature restriction is only really necessary if if suppression pool level reaches 26 feet. That is, suppression pool level reaches 26 feet and pool temperature is less than 140 F, then it is necessary for the operator to manually transfer HPCI suction to the suppression pool in order to mitigate the rise in pool level and avoid potential malfunction of the HPCI system. Since it is preferable for HPCI to take suction from the CST because the water is demineralized, the EOPs are revised as part of the change to remove the 140 F suppression pool temperature restriction for suppression pool levels less than 26 feet.
The following components are affected by the proposed change:
~ HPCI system (suction valves HV-155/255-F042 and HV-155/255-F004, HPCI pump 1P204, 1P209 /2P204, 2P209, and HPCI turbine 1S211/2S211),
~ RCIC turbine (1S212/2S212),
Primary Containment and Reactor Building safety-related systems, structures, and components affected by LOCA/SRV hydrodynamic loads, 250V DC Control Center 1D264 (2D264),
s HPCI Relay Panel 1C620/2C620, and
~ Safety related valves on piping connected to suppression chamber.
The following is a discussion of the safety functions of each of the components listed above and consideration and resolution of all potential effects on those functions as a result of the proposed change:
n n n
- HPCI Suction Valve HV-155/255-F042 has two safety functions: suction transfer and containment isolation. The safety function of HPCI Suction Valve HV-155/255-F004 is suction transfer. These functions are discussed below.
ENCLOSURE A TO PLA-4428 Page 3 of 11 In an accident situation, HPCI initially injects water from the CST, but the pump suction automatically trarisfers to the suppression pool on high pool level or low CST level (FSAR, Section 6.3.2.2.1) The transfer is accomplished by the automatic opening of valve HV-155/255-
~
F042 (HPCI suppression pool suction valve) and automatic closing of valve HV-155/255-F004 (HPCI CST Suction valve). This change only affects the open/close signal which is generated by high suppression pool level.
The safety function of the suction transfer logic relative to high suppression pool level is discussed in the HPCI Design Basis Document (DBD 004), Section 2.16.3.3.1 as follows:
"The basis for the suction transfer on high suppression pool level is to prevent the HPCI System from contributing to the further increase in the suppression pool level.
The maximum suppression pool water level is dictated by the need to maintain sufficient air space to accommodate the non-condensable gases that are blown down to the suppression chamber during an accident. If the suppression pool water level was too high, the non-condensable gases would cause the containment pressure to exceed design values. The water level would also be a factor in the calculation of pool swell loads which would arise from the gaseous discharge from the containment drywell to the wetwell during the early stages of a postulated Design Basis Accident, and from the blowdown loads generated by an ADS depressurization event. A small break LOCA with HPCI injection may raise suppression pool level to 24 ft. The design basis for the hydrodynamic loads due to SRV/ADS blowdown are based on a maximum 24 A. pool level. Exceeding 24 ft. has the potential to produce SRV loads that may exceed the suppression pool design basis."
In summary, the HPCI DBD addresses three concerns with regard to suppression pool water level:
~ HPCI operation contributing to an initial pool level > 24't the time a DBA occurs.
~ HPCI operation causing pool level to exceed 24'uring the course of a Design Basis LOCA.
~ HPCI operation causing pool level to increase above 24'uring a small break LOCA which subsequently requires initiation of ADS.
This Safety Assessment addresses each of these three concerns with regard to the proposed change.
HPCI suction valve HV-155/255-F042 is also a containment isolation valve (Technical Specification Table 3.6.3-1). This valve must close within 90 seconds upon receiving any of the isolation signals listed in Technical Specification Table 3.3.2-1 ~
ENCLOSURE A TO PLA-4428 Page 4 of 11 42k 4 te i The proposed change may affect the manual and automatic suction transfer capability for HPCI.
With the proposed change, HPCI will take suction from the CST until the water source is depleted (auto transfer), or until reactor pressure drops below the shutoff head of low-pressure ECCS, or until pool level approaches 26 feet (manual transfer), depending on the accident conditions. Under current plant design, HPCI suction auto swaps to the pool when level reaches 23'-9". Operability analysis2 for susceptibility to thermally-induced pressure locking shows that the HPCI suppression pool suction valve F042 will open to allow the HPCI auto suction swap to the pool when level reaches 23'-9". The auto transfer capability remains effective because pool temperature does not increase substantially by the time level reaches 23'-9". With the proposed change, however, suction transfer to the suppression pool may be required later in the accident if pool level reaches 26 feet (proposed EOP requirement) or ifwater is depleted from the CST. In this situation, the pool temperature will have increased significantly. An analysis has been performed in accordance with the guidance of Generic Letter 95-07 for F042 and determined it not to be susceptible to pressure locking under the analyzed conditions.
As discussed above, the proposed change may alter the time at which the F042 valve opens and F004 closes during a design basis event. HPCI suction is maintained on the CST until it transfers to the suppression pool on low CST level or is manually transferred by the operator. Voltage requirements will be unaffected as discussed below, and this change has no effect on the containment isolation logic of F042.
-ThHPCtp p d ti ppli I <<h li if clad temperatures in the event of a small break LOCA which does not result in rapid depressurization of the reactor vessel.
2gmp - Eliminating the HPCI suction transfer on high suppression pool level will increase HPCI reliability in accidents involving elevated suppression pool temperature. Since the HPCI lube oil is cooled by the pumped fluid, failure of the pump from overheating is precluded if suction is maintained on the CST. This change has no other effects on the HPCI pump.
3mhing - There are four potential problems which have been identified with operating, tripping, and restarting of the HPCI turbine with high suppression pool level. These potential problems are listed below along with their resolutions.
Generic Letter 95-07 SSES Operability Assessment, "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves," 11-15-95.
ENCLOSURE A TO PLA-4428 Page 5 of 11
~~~ The potential was identified for HPCI turbine exhaust line flooding in a small-break accident ifHPCI trips with pool level above the exhaust line containment penetration (25.6 feet above the bottom of the pool). It was postulated that water would leak through turbine exhaust line check valve F049, and a water-hammer would then occur upon restart of the HPCI turbine possibly damaging the turbine and associated piping.
ll J+~l~i Based on expected leakage rates through the F049 valve, it was concluded that leakage will be contained well within the turbine exhaust line drain pot. The study shows that even ifthe initial drain pot level is at the high-level alarm set point (75% full),
there is sufficient capacity to allow for a leakage rate which is 50 times the measured value. Therefore, a water hammer will not occur upon restart of the turbine.
~m With the proposed change, water level, in a small break accident, may reach 27.2 feet and completely submerge the horizontal section of the turbine exhaust line which penetrates the containment.> IfHPCI trips with pool level > 27.2 feet, water will flood the horizontal section of piping up to isolation check valve F049. When this occurs, the column length of water in the exhaust line increases by about 25 feet. Due to inertial effects, a higher turbine exhaust pressure will develop as this column of water is expelled upon auto restart of the turbine. This raises a potential that the HPCI pressure-relief diaphragms will rupture upon turbine restart and render the system inoperable. OP-152/252-001 instructs the operator to start HPCI in accordance with EO-100/200-032 ifsuppression pool level > 26 feet. Under conditions of high pool level, EO-100/200-032 requires the operator to start the HPCI turbine with turbine flow control in manual and speed set to minimum. Speed is increased slowly to gradually clear the exhaust line, and the operator monitors turbine exhaust pressure during this evolution. Although OP-152/252-001 addresses HPCI restart with high pool level, there is no guarantee that this procedure will result in successful restart of the system. The restart procedure has never been used, and it cannot be tested. In addition, the restart evolution is too complicated to be analyzed with a reasonable degree of uncertainty.
Jhsg3zLtIILip The EOPs will be modified to include operator action to manually transfer HPCI suction from the CST to the suppression pool if pool level reaches 26 feet with pool temperature less than 140 'F. If HPCI trips with pool level at 26 feet, the water depth within the horizontal section of exhaust piping (20 inch pipe) up to isolation check valve F049 will be only 5 inches. This represents a minor reduction in steam flow area, and there would be no adverse effects upon restart of the HPCI turbine. In a small break accident, pool level can reach 26 feet only for a narrow range of break sizes. Moreover, the operator action to manually transfer HPCI suction would only be required in the long-term part of the accident (about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with pool level initially at the nominal level of 23 feet). Ifpool temperature is greater than 140 'F when pool level reaches 26 feet, HPCI 25.6 feet corresponds to the bottom of the horizontal piping which penetrates the containment; the top of this piping is at an elevation of 27.2 feet (piping ID is 19.2 inches).
ENCLOSURE A TO PLA-4428 Page6of11 suction will not be transferred to the pool because adequate cooling of the HPCI pump cannot be assured. However, ifthe pump suction auto transfers to the suppression pool on low condensate storage tank level when suppression pool temperature is greater than 140 F, the operator will continue to use HPCI as necessary. With HPCI injecting to the vessel, suppression pool temperature is expected to exceed 140 F only for beyond-design-basis events. It is appropriate to continue using HPCI with suction temperature greater than 140 F in a beyond-design-basis event because the system may be required to prevent actuation of ADS, or it may be required to prevent core damage in accident scenarios where low pressure injection systems or the depressurization capability are unavailable.
~~e~ If, in a small-break accident, suppression pool level reaches 28.5 feet, the air intake for the HPCI turbine exhaust-line vacuum breakers (F076 and F077 on the HPCI turbine exhaust line) becomes submerged. The most serious consequence of disabling the vacuum breakers is the potential for water hammer on the turbine exhaust-line check valve (F049) in the event of a system trip.
Q~~ii This problem is eliminated by the resolution to Problem 2 above. Note that ifpool temperature is greater than 140 'F, there is no point in manually transferring HPCI suction back to the pool under any circumstance because continued operation of the system cannot be assured. Suppression pool temperature is not expected to exceed 140 'F in a design-basis small break accident while HPCI is operating.
QgMcrg 4 The HPCI turbine exhaust pressure for continuous steady-state operation may exceed the design limit.
Rggggfign, The HPCI turbine is designed to operate at a maximum continuous exhaust pressure of 65 psia (HPCI DBD004, Requirement 2.3.3.1.4). Analysis shows that there is margin to the design exhaust pressure limit of 65 psia. 'mple Note that suppression pool level will not reach 26 feet for all small breaks. For example, in the case of a 1" line break, pool level only increases by about 4 inches.
In events where suppression pool level does reach 26 feet, level does not remain at this elevation because suppression pool letdown begins to reduce pool level once HPCI suction is manually transferred to the suppression pool.
The operator can monitor suppression pool water level with control room level indicators LI-157775A/25775A and LI-15775B/25775B. The range of these safety-related instruments is 18-26.5 feet. These indicators are located on panel 1C601/2C601.
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ENCLOSURE A TO PLA-4428 Page 7 of 11 r ine In accordance with its design basis, the RCIC system functions to maintain sufficient water inventory in the reactor to permit adequate core cooling following a reactor vessel isolation event accompanied by loss of feedwater. RCIC does not perform a safety-related function, except for containment isolation. The RCIC system is, however, classified as an Appendix R Safe Shutdown System and may be used for vessel injection in the event of a fire on site.
As discussed above, RCIC is used to provide coolant makeup following a reactor vessel isolation event. In a MSIV closure event, HPCI and RCIC would initiate on low RPV water level. These systems would inject to the vessel until they automatically trip when level reaches 54 inches.
There would be no additional HPCI/RCIC initiations within the first 10 minutes of the event.
After 10 minutes, it can be assumed that the operator will use RCIC for RPV makeup, and HPCI will be used for pressure control (CST-to-CST mode). Therefore, the long-term part of the scenario (t > 10 minutes) is completely unaffected by the proposed change.
RCIC may be used for Appendix R Safe Shutdown; however, the shutdown scenario assumes vessel isolation and the effects are similar to the isolation event described above.
If the initial suppression pool level is at the Technical Specification limit of 24 feet, then suppression pool level will be about 2 inches higher with the proposed change.4 This small level change has negligible effect on RCIC turbine exhaust pressure, and a large margin remains to the turbine exhaust line elevation of 26 feet. In the more realistic situation where the initial SP level is at a nominal value of 23 feet, the proposed change has no effect on containment response.
Although RCIC is not designed for vessel makeup in a small break accident, it is prudent to examine the impact of the proposed change on RCIC operation under LOCA conditions. The issues and resolutions for RCIC turbine operation with, elevated suppression pool level are essentially the same as those presented above for the HPCI turbine. Problem 1 is not an issue for the RCIC turbine, however, because the bottom of the horizontal section of turbine exhaust piping corresponds to 26 feet, and therefore the horizontal run of piping cannot become flooded because pool level will not exceed 26 feet in a design basis accident. Also, Problem 4 is not applicable to the RCIC system as its maximum turbine exhaust pressure for continuous operation is only 25 psia (DBD041, Rev. 0, Requirement 2.3.2.1.4), and this value would be exceeded in a small break accident even if there is no increase in suppression pool level.5 Moreover, the maximum increase in suppression pool level is two feet (24 feet to 26 feet) which has negligible With current plant design, HPCI would take suction from the suppression pool ifthe initial pool level > 23'-9",
and the volume of coolant injected to the RPV prior to the HPCI trip on 54" corresponds to a 2" level decrease in the suppression pool.
Peak wetwell atmosphere pressure is calculated to be about 35 psia for a small break accident.
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ENCLOSURE A TO PLA-4428 Page 8 of 11 effect on RCIC turbine operation as it corresponds to a pressure increase of only 0.86 psi at the turbine exhaust.
'm. nain etad eac r u'n fe - ateI t t icie an cte I a c LOCA and SRV hydrodynamic loads potentially affect all primary containment safety-related structures, systems, and components. Dynamic loads on the primary containment indirectly affect the reactor building, and all safety-related equipment in the reactor building, because the structures are interconnected. The primary containment and reactor building safety-related structures, systems, and components perform numerous safety functions. These safety functions are described in the FSAR.
As discussed above, the purpose of the HPCI suction transfer on high pool level is to prevent containment hydrodynamic loads associated with a LOCA/ADS blowdown from exceeding design limits. Removal of the HPCI auto transfer on high SP level will allow slightly higher pool levels during plant accidents and transients; however, it has been concluded that the higher pool level does not lead to violation of design limits for hydrodynamic loads.
Therefore, the proposed change has no adverse effects, with respect to LOCA/SRV hydrodynamic loads, on primary containment or reactor building safety-related structures, systems, or components.
Note that the higher suppression pool water level associated with the proposed modification does not impact the diaphragm-slab-differential-pressure or drywell-negative-pressure analyses (FSAR Section 6.2.1.1.4). These analyses assume that drywell sprays are initiated during a small break accident and that all noncondensible gases are contained within the wetwell air space at the time of the spray actuation. In addition, the wetwell temperature is non-mechanistically set to 50 F. With the HPCI auto suction transfer elimination, SP level can rise to 26 feet in a small break accident. This causes a reduction in the wetwell air space volume. Ifa smaller wetwell air space volume is considered in the diaphragm-slab-differential-pressure and drywell-negative-pressure analyses, the results will be more favorable because the wetwell will exhibit a faster pressure response upon opening of the vacuum breakers. That is, the wetwell pressure will more closely follow the drywell pressure. Therefore, it is conservative to neglect the reduction in suppression chamber free volume when computing the diaphragm slab differential pressure and tlie drywell'peak negative pressure.
te 250V DC Control Center 1D264 (2D264) is a Division II safety related power source providing power to HPCI DC motor operated valves and the RCIC FOS4 vacuum breaker valve.
ENCLOSURE A TO PLA-4428 Page 9 of 11 The 250 VDC power system as described in FSAR Section 8.3.2.1.1.2 will have a decrease in load due to the removal of the auxiliary relay from panel 1D264 (2D264). The change in battery load and in line voltage drop are negligible. The changes are documented in revisions of the applicable calculations. Dynamic qualification of the panel was evaluated and is not adversely affected by the removal of the relay.
A consideration in formulating the battery load profile is the time line associated with the addition of various loads on the batteries. For the F042 valve, however, it cannot be determined when a low CST tank level may occur because the CST is not seismically qualified. Therefore, the design-basis evaluation for the battery loads assumes that the F042 valve starts to open and F004 starts to close during the heaviest loaded segment of the 1D660 (2D660) Battery loading sequence. As a result, no changes are required to the 1D660 (2D660) Battery load profile because of the proposed change.
1 encl 2 62 HPCI Relay Panel 1C620 (2C620) provides Division II control and instrument functions for the HPCI system. Safety related 125V DC and 120V AC distribution panels provide power for the circuit components in this relay panel.
The 125 VDC power system as described in FSAR Section 8.3.2.1.1.1 will have a decrease in load due to the replacement of the K19 relay in panel 1C620 (2C620). The change in battery load and change in line voltage drop are negligible. The change in load and voltage drop is documented in revisions of the applicable calculations. The replacement relay and its mounting are dynamically qualified and the dynamic qualification of the panel is not adversely affected by this replacement.
i 'n nn t t u Safety-related valves on piping connected to the suppression chamber provide flow paths for ECCS and SP cooling. Other valves on piping connected to the suppression chamber include SRVs and vacuum breakers on the downcomer vents. SRVs prevent overpressurization of the reactor vessel, and the downcomer-vent vacuum breakers equalize pressure across the drywell floor in the event of a LOCA.
MOVs - Suppression pool level could potentially increase by 2 feet during a design basis small break accident, as a result of this modification (i.e., 24 feet to 26 feet). The increase in pressure
ENCLOSURE A TO PLA-4428 Page 10 of 11 due to the additional 2 feet is 0.86 psi (assuming suppression pool temperature of 90 F). This small increase in pressure will not adversely affect the pressure retaining capability of any valve on piping connected to the suppression pool. In addition, the hP across these valves could increase by 0.86 psi, depending on valve function. The ability to open or close these valves in accordance with applicable design criteria is not affected by this change in hP as documented in MOV design-basis calculations. These calculations conclude that there are no adverse effects to the operation or performance of any valve on piping connected to the suppression pool as a result of this small pressure increase.
The HPCI suction transfer logic elimination does not increase the severity of the suppression pool temperature transient in a small break accident. In fact, the suppression pool temperature rise (for a small-break LOCA) would be larger under the current plant configuration than it would be with the proposed modification installed. A smaller suppression pool temperature rise would result because of the additional mass added to the suppression pool when HPCI suction is maintained on the condensate storage tank for the duration of the accident. Since the energy deposited in the suppression pool is unchanged by the proposed modification, the additional mass leads to a smaller increase in pool temperature. Thermal locking effects due to suppression pool temperature increase are already considered in the Generic Letter 95-07 operability evaluation, and since the suppression pool temperature response for the proposed modification is bounded by the current response, there is no impact on valve thermal locking other than that already discussed for the HPCI suppression pool suction valve (F042).
Vacuum Breakers - Allowing suppression pool level to potentially increase to 26 feet in a design-basis accident does not impact operation of downcomer-vent vacuum breakers because the vacuum breakers are located 42 feet above the bottom of the suppression pool.
SRVs/Tailpipes - The increased suppression pool level associated with the proposed change has no effect on SRV operation because flow through the SRVs is choked. SRV flow is decoupled from downstream conditions when the flow through the valves is choked.
The higher suppression pool level does, however, lead to a higher peak pressure within the SRV tailpipe upon valve actuation. When a SRV opens, the SRV tailpipe rapidly pressurizes as the slug of water within the pipe is expelled. As suppression pool water level increases, there is an equivalent increase in the water column height within the tailpipe, and consequently, the maximum pressure buildup within the tailpipe increases with pool level. Although higher suppression pool levels are expected under accident conditions with the proposed change, the magnitudes are such that there is no threat of SRV tailpipe failure. With the proposed change, suppression pool level could potentially increase to 26 feet. Design basis loads on the SRV system are conservatively based on an initial pool level of 35 feet. This level is nine feet above the maximum suppression pool level which could occur with the proposed change. Actually, for a loss-of-coolant accident, the available margin is much greater than nine feet because reactor pressure continually decreases during the event. For beyond-design-basis conditions, the EOPs
ENCLOSURE A TO PLA-4428 Page 11 of 11 require reactor,depressurization before pool level reaches the point where SRV tailpipe integrity is threatened.
im in in The proposed change was evaluated for impacts on containment hydrodynamic loads for LOCAs, plant transients, and special events (ATWS and SBO). Also evaluated was HPCI ope'ration with elevated suppression pool level. The impact on containment loads was examined qualitatively by determining the magnitude of suppression pool level increase in accidents and plant transients which involve HPCI operation.
The impact of high suppression pool level on the HPCI turbine was evaluated including water hammer concerns and the effects of increased pool level on turbine exhaust pressure.
Suppression pool letdown capability during a design-basis accident was evaluated and found acceptable.
Changes in 125 VDC and 250 VDC battery loading and the negligible impact on voltage drop associated with the applicable DC distribution systems were evaluated and found acceptable for:
~ 125 VDC 1D620 Battery Load Profile
~ 125 VDC 2D620 Battery Load Profile
~ 125 VDC 1D62401 Voltage Drop 250 VDC Voltage Drop
~ 250 VDC 1D660 (2D660) Battery Load Profiles
~ 250 VDC 1D264061 Voltage Drop
~ 250 VDC 2D264061 Voltage Drop
ENCLOSURE B TO PLA-4428 Page 1 of9 HPCI PUMP SUCTION AUTOMATICTRANSFER TO SUPPRESSION POOL LOGIC ELIMINATION(UNITS 1 AND 2)
Pennsylvania Power & Light Company has evaluated the proposed Technical Specification change in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration. The criteria and conclusions of our evaluation are presented below.
- 1. The proposed change does not involve an increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated.
Based on the following discussion for the containment, reactor building, HPCI & RCIC systems, and the safety-related valves in piping connected to the suppression pool, the proposed action does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety, as previously evaluated in the SAR.
As discussed in the Safety Assessment for this change, elimination of the HPCI auto suction transfer on high suppression pool level will allow higher suppression pool water levels in accidents" and transients which involve HPCI operation. The impact of the higher suppression pool levels were examined for the following design-basis accidents and transients:
~ Loss of Coolant Accidents inside containment (FSAR $ 6.2.1.1.3.3),
~ Inadvertent Safety/Relief valve opening (FSAR $ 15.1.4),
~ Primary system break outside containment (FSAR $ 3.6A),
, ~ Inadvertent HPCI initiation (FSAR $ 15.5. 1),
~ Loss of feedwater flow (FSAR $ 15.2.7),
~ Loss of Offsite AC Power (FSAR $ 15.2.6),
~ Loss of Main Condenser vacuum (FSAR $ 15.2.5),
~ Inadvertent MSIV closure (FSAR $ 15.2.4),
~ Turbine trip (with and without bypass) (FSAR $ 15.2.3),
~ Generator Load Rejection (with and without bypass), (FSAR $ 15.2.2), and
~ Pressure regulator failure-closed/open (FSAR $ 15.2.1 & 15.1.3).
These accidents and transients were selected for evaluation because they involve an initiation of the HPCI system either inadvertently or as a result of a decrease in vessel inventory and/or coolant level. Two special events, ATWS and SBO, are also considered along with the design basis events listed above.
ENCLOSURE B TO PLA-4428 Page20f9 It was concluded that design-basis SRV and LOCA loads envelop the loads expected with the proposed change. Therefore, the proposed change does not increase the failure probability of any primary containment or reactor building structure, system or component which is affected by LOCA/SRV hydrodynamic loads. The major findings which lead to this conclusion about SRV and LOCA loads are summarized below:
DBA dynamic pressure loads are based on a maximum initial suppression pool level of 24 feet. The proposed modification to the HPCI suction transfer logic does not affect the initial pool level or the initial suppression chamber air space volume. During normal plant operation, suppression pool level (and hence suppression chamber air space volume) is controlled by Technical Specification requirements.
It I
For LOCAs other than the DBA, the containment is designed for ADS blowdown loads in combination with the LOCA loads. For an intermediate break, the proposed HPCI modification does allow suppression pool level to exceed 24 feet by a small amount. ADS loads are, however, independent of suppression pool level when the downcomer vents are cleared. Therefore, the proposed modification has no influence on ADS hydrodynamic loads for an intermediate break.
~ For small breaks, HPCI injection prevents ADS actuation. Nevertheless, SRV actuations occur during the RPV cooldown. Downcomer vents are opened in the beginning part of the accident, but close later on as the break enthalpy decreases. When the downcomer vents are cleared, the level inside the SRV tailpipe is not influenced by pool level, and therefore, the SRV hydrodynamic loads are unaffected by the proposed modification. During the phase of the accident in which the downcomer vents are sealed with water, there are no wetwell LOCA hydrodynamic loads, but the SRV loads are dependent on SP water level. In this case, SRV loads are acceptable because SP water level is always below the Load Limit curve.
~ ADS actuation would be required in the event of a HPCI failure during a small-break accident. IfHPCI fails during the phase of the accident in which the downcomer vents are cleared, then ADS loads would be acceptable because water level (and air volume) within the SRV tailpipes is independent of pool level. Even ifHPCI failure occurs in the latter part of the accident where the downcomer vents are sealed, ADS loads are acceptable because water level is always well below the Load Limit curve.
~ Under non-LOCA conditions, the containment is designed for simultaneous actuation of all 16 SRVs. The Load Limit Line defines the acceptable operating region, in terms of reactor pressure and suppression pool level, for actuation of all 16 SRVs. Following a plant transient involving HPCI operation, the suppression pool level is always below the Load Limit curve, and only a small number of SRVs actuate to remove decay heat from the reactor.
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ENCLOSURE B TO PLA-4428 Page 3 of 9 The proposed change does not increase the probability of an equipment malfunction in the HPCI system. In fact, the change eliminates the potential failure of the HPCI suction auto-transfer on high suppression pool level since that logic is removed. Potential spurious auto-transfer associated with high suppression pool logic is also eliminated. HPCI suction auto-transfer on low CST level and its potential to fail are unchanged by this change. Also, the change does not affect the manual suction transfer from the CST to the suppression pool.
As discussed in the safety assessment for this change, the proposed change has no adverse effects on HPCI valves, pump, or turbine. Therefore, elimination of the HPCI suction auto transfer logic (on high suppression pool level) does not increase the probability of a HPCI malfunction. The consequence of a HPCI failure in a design-basis accident is evaluated in NEDC-32071P Rev.l, "Susquehanna Steam Electric Station Units 1 and 2 SAFER/GESTR-LOCA Loss-of -Coolant Accident Analysis." With regard to the fuel, the consequence of a HPCI failure is unaffected by the proposed change.
If HPCI fails in a design-basis small break accident, ADS actuation would be required. ADS loads continue to be enveloped by design loads with the proposed change. Therefore, the proposed change does not increase the consequences of a HPCI failure.
On a component level, the failure probability and consequences of failure associated with the AX relay in 250 VDC Control Center 1D264 (2D264) are eliminated because the relay is disconnected and removed by this modification. Since the control functions of K19 in panel 1C620 (2C620) have been eliminated, the failure of the relay has no effect on HPCI suction valve F042 operation.
The 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 1C620 (2C620) both receive power from battery systems during Station Blackout. Removal of the relay from 250 VDC Control Center 1D264 (2D264) and the replacement of the relay in HPCI Relay Panel 1C620 (2C620) decreases the load on the battery systems by a small amount. The change in battery load and line voltage drop is negligible and is documented in applicable calculations.
Dynamic qualification of the subject equipment is not adversely affected by this modification as documented in applicable calculations.
As discussed in the safety assessment for this change, RCIC is used to provide coolant makeup following a reactor vessel isolation and for an Appendix R shutdown scenario. The Appendix R event also assumes the reactor vessel is isolated. These events are discussed in Section 15.2.4 of the FSAR and in the FPRR. The proposed change has no adverse effects on RCIC turbine operation following a MSIV closure (see discussion in the safety assessment for this change).
ENCLOSURE B TO PLA-442S Page4of9 Therefore, there is no increase in the RCIC failure probability for the MSIV-closure event or the Appendix R shutdown scenario. The consequence of RCIC failure is unchanged by the proposed if modification; RCIC fails, HPCI is available as a backup system.<
Although RCIC is not designed for mitigation of a small break accident, the effect of the proposed change on RCIC turbine operation for such an accident was evaluated in the safety assessment for this change. The assessment concludes that the proposed change has no adverse effects on RCIC operation, and therefore, there is no increase in RCIC failure probability during a small break accident. Failure of RCIC in a small break accident would require ADS initiation only for a particular break flow which is slightly greater than HPCI injection capability. But ADS initiation has already been considered when evaluating the consequences of HPCI failure during a small break accident.
MOVs - The proposed change could potentially lead to a maximum suppression pool level of 26 feet in a design-basis accident. This is 2 feet above the maximum design level of 24 feet. As discussed in the safety assessment for this change, this is equivalent to a pressure increase of 0.86 psi at the bottom of the suppression pool. This small pressure increase has negligible effect on valve operation, and therefore, there is no increase in the probability of a failure or malfunction of valves in piping connected to the suppression pool.
Vacuum Breakers - Allowing suppression pool level to potentially increase to 26 feet in a design-basis accident does not affect the failure probability of downcomer-vent vacuum breakers because the level is well below the vacuum breaker elevation of 42 feet.
SRVs/Tailpipes - As discussed in the safety assessment for this change, the increased suppression pool level associated with the proposed change does not have any adverse effect on SRV operation or on the structural integrity of the SRV tailpipe.
- 2. The Proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
The following discussion concerning the impact of the change on the primary containment, the reactor building, the HPCI system, and safety-related valves, provides the basis for this conclusion.
The HPCI suction transfer logic is not necessary to maintain LOCA loads within design limits because these dynamic pressure loads are characterized in terms of the SP level at the initiation DBD041, Rev. 0, p. l.
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ENCLOSURE B TO PLA-4428 Page 5 of9 l
of the accident. That is, LOCA blowdown tests were conducted without the removal of water from the suppression chamber section of the test tank.7 The increase in pool level realized
,during these tests was proto-typical of the pool level increase expected at Susquehanna.
Removal of the HPCI suction transfer logic on high pool level does not affect suppression pool level at the initiation of a DBA.8 In addition, the HPCI suction transfer logic is not necessary to maintain SRV/ADS blowdown loads within design limits. SRV dynamic pressure loads consist of two components: air clearing loads and steam condensation loads. The steam condensation loads are bounded by the more severe air clearing loads which are caused by gas bubble oscillations following the expulsion of noncondensible gas from the SRV tailpipe. Air clearing loads are a function of reactor pressure and water level inside the SRV tailpipe.
Depending on the break size and location, the downcomer'vents may be cleared for the entire time that HPCI is operating, or they may reseal in the latter part of the accident. When the downcomer vents are cleared, the level inside the SRV tailpipe is depressed to the elevation coinciding with the bottom of the downcomer pipes, and it is therefore decoupled from the rising suppression pool level. In this situation SRV air-clearing loads are unaffected by the proposed change.
When the downcomer vents are sealed with water, the Load Limit line can be used to determine if SRV/ADS loads are enveloped by design loads. For the most limiting event, which is the small break LOCA, the overall safety margin increases as pool level rises during the event. This is because the decrease in reactor pressure more than offsets the adverse effects associated with the rise in pool level.
Since LOCA and SRV dynamic loads remain bounded by design loads, dynamic loading of primary containment and reactor building structures, systems, and components are unaffected by the proposed change. Therefore, with respect to dynamic loads, the proposed change does not create the possibility for an accident or malfunction of a different type than any evaluated in the SAR.
There are no new HPCI turbine failure modes introduced by the higher suppression pool levels which can occur with the proposed change. Turbine exhaust pressure remains well below the design limit of 65,psia. In addition, the higher pool level does not create the possibility of water hammer damage to the turbine discharge piping. Ifthe operator fails to control RPV level less than +54" (single operator error) in the long-term part of the small-break accident when suppression pool level is > 25.6 feet, leakage through check valve F049 is such that it will be 7 SSES DAR, Section 9.4.1 Suppression pool level must be maintained less than 24 feet in accordance with Technical Specification 3.6.2.1.a.
ENCLOSURE B TO PLA-4428 Page 6 of9 contained well within the volume of the turbine-discharge-line drain pot. Note that suppression pool level is limited to 26 feet by operator action. Furthermore, suppression pool level can reach 26 feet only for a particular range of small breaks, and for this range of small breaks, suppression pool level would exceed 25.6 feet for only -10 minutes of the accident duration. This corresponds to about 10% of the time that HPCI is operating. Thus it is very unlikely that HPCI would trip with pool level > 25.6 feet.
If check valve F049 is failed during the small-break accident (single equipment failure), the turbine exhaust line would become flooded if the HPCI system tripped during the 10 minute interval when suppression pool level > 26 feet; however, it is not necessary to postulate an operator error (failure to control RPV level <+54") along with the check valve failure. A small break accident with failure of check valve F049 and failure of the operator to control RPV level as required by the EOPs, in a narrow time interval during the long-term part of the accident, is beyond the plant design basis.
A new type of malfunction does not occur even in the beyond-design-basis condition where failure of check valve F049 is considered along with failure of the operator to control RPV level
< 54" in the narrow time interval when pool level is > 25.6. With these failures, the turbine exhaust piping will become flooded, and the system may fail on restart. The General Electric Company has performed an analysis to determine the consequences of a HPCI start with flooding of the turbine and adjacent exhaust line 9 The analysis, which addresses a potential design deficiency in the HPCI barometric condenser, shows that the containment penetration head fitting and interface piping will not fail as a result of the water hammer associated with the HPCI start. Since failure of the HPCI system is already considered in the plant design-basis accident analysis; this is not a different type of malfunction than that already considered.
No new failure modes are introduced by the hardware changes in the 250 VDC Control Center 1D264 (2D264) and HPCI Relay Panel 1C620 (2C620). Some failure modes are eliminated by the proposed change. Specifically, the potential failure of the HPCI suction auto-transfer on high suppression pool level is eliminated since that logic is removed. Potential spurious auto-transfer associated with high suppression pool logic is also eliminated. HPCI suction auto-transfer on low CST level and its potential to fail are unchanged by this change.
On a component level, potential failure modes for the AX relay in 250 VDC Control Center 1D264 (2D264) are eliminated by this modification because the relay is disconnected and removed by this change. The potential failure modes for the relay K19 in panel 1C620 (2C620) are unchanged. Since the control functions of K19 have been eliminated, the failure of the relay has no effect on HPCI suction valve F042 operation.
GKR-93-001, "NRC and Utility Notification of Closeout of GE PRC92-05, Potential Design Deficiency on HPCI," January 6, 1993.
ENCLOSURE B TO PLA-442S Page7of9
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Removal of the relay from 250 VDC Control Center 1D264 (2D264) and the replacement of the relay in the HPCI Relay Panel 1C620 (2C620) changes the load on the battery systems by a small amount. The change in battery load and change in line voltage drop are negligible and they do not adversely affect the performance of the panels or battery systems. In addition, seismic qualification of the panels is not adversely affected by this change.
As discussed in the safety assessment for this change, the proposed change has no adverse effects on RCIC turbine operation. Therefore, the proposed change cannot result in a new RCIC failure mode.
MOVs - The increased suppression pool water level which can occur as a result of the proposed change does not create a failure mechanism for safety-related valves on piping connected to the suppression pool. The pressure differential for any valve on piping connected to the suppression pool will increase by at most 0.86 psi. This change in differential pressure has negligible effect on valve operation.
Vacuum Breakers - The proposed change cannot lead to malfunction of the downcomer-vent vacuum breakers as the maximum level expected in a design-basis event is 26 feet, and the vacuum breakers are located at 42 feet above the suppression pool floor.
SRVs/Tailpipes -. There is no interaction between increased suppression pool level and SRV operation since the flow through the SRVs is ch'oked and therefore decoupled from downstream conditions. Also, the increased suppression pool level cannot lead to failure of the SRV tailpipe because the potential level increase is well below the SRV Tailpipe Level Limit.10 If suppression pool water level is below this limit, there is no concern of tailpipe failure due to overpressurization. The minimum value of the SRV Tailpipe Level Limit is 35 feet.>1 This is 9 feet above the maximum level expected in a design-basis accident. For beyond-design-basis events, SRV tailpipe integrity is protected by the BOP requirement to depressurize the reactor on the SRV Tailpipe Level Limit.
- 3. The proposed change does not involve a significant reduction in a margin of safety.
The HPCI Technical Specifications ensure that the system is capable of providing adequate core cooling to limit clad temperatures in the event of a small break LOCA which does not result in This limit is defined in EO-100/200-103.
Bechtel Calculations PUP-15598-S2 & PUP-15598-SG, and PLE-15315 (March 2, 1992).
I ENCLOSURE B TO PLA-4428 Page 8 of 9 rapid depressurization of the RPV (Technical Specification Section 3/4.5.1 &, 3/4.5.2). The proposed change has no adverse affects on the injection capability of the,HPCI system.
Therefore, the safety function of the system is not degraded, and there is no reduction in the margin of safety as defined in the basis for the HPCI Technical Specifications.
r n Removal of the HPCI auto suction transfer on high suppression pool level does not affect the Technical Specification requirement to maintain suppression pool water level between 22 and 24 feet (Technical Specification 3.6.2.1). Therefore, the maximum containment pressure during the design-basis accident is unaffected by the proposed change, and there can be no reduction in the margin of safety as defined in the basis for Technical Specification 3.6.2.1. Furthermore, a detailed examination of the reactor and containment response under accident and transient conditions involving HPCI operation found no situations where the auto suction transfer was necessary to maintain LOCA and SRV loads within the design basis envelope. Therefore, from the standpoint of LOCA/SRV hydrodynamic loads, the proposed change does not reduce the margin of safety for any primary containment or reactor building structure, system, or component.
The basis for Technical Specification 3.7.3 states that the RCIC system is provided to assure adequate core cooling in the event of a reactor isolation with loss of feedwater flow. The proposed'change does not prohibit RCIC from performing this function, nor does it degrade in any way the core cooling capability of RCIC. Therefore, there is no reduction in the margin of safety as defined in the basis for Technical Specification 3.7.3.
ve MOVs - The increase in suppression pool water level which can occur as a result of the proposed change does not reduce the margin of safety for safety-related valves on piping connected to the suppression pool. The pressure differential for any valve on piping connected to the suppression pool will increase by at most 0.86 psi. This change in differential pressure has negligible effect on valve operation.
Vacuum Breakers - The proposed change cannot reduce the margin of safety as discussed in the basis for Technical Specification 3.6.4 because the maximum level expected in a design-basis event is 26 feet which is well below the downcomer-vent vacuum breaker elevation of 42 feet.
SRVs/Tailpipes - There is no interaction between increased suppression pool level and SRV operation since the flow through the SRVs is choked and therefore decoupled from downstream conditions. Consequently, there is no reduction in the margin of safety as defined in the bases for Technical Specifications 3.4.2 (safety valve function) and 3.5.1.d (ADS function). Also, the
1 II ENCLOSURE B TO PLA-4428 Page 9 of9 incr'eased suppression pool level does not lead to a reduction in the margin of safety for the SRV tailpipes because the tailpipes can operate safely with pool levels up to 35 feet. This is nine feet above the maximum suppression pool level that can occur in a design-basis accident with the proposed change. For beyond-design-basis events, SRV tailpipe integrity is protected by the EOP requirement to depressurize the reactor on the SRV Tailpipe Level Limit.12 As discussed previously, removal of the relay from 250 VDC Control Center 1D264 (2D264) and the replacement of the relay in the HPCI Relay Panel 1C620 (2C620) changes the load on the battery systems by a small amount. The change in battery load and change in line voltage drop are negligible and therefore they do not reduce the margin of safety for the panels or battery systems. In addition, seismic qualification of the panels is not adversely affected by this change so there is no reduction in the margin of safety for seismic events.
An environmental assessment is not required for the proposed change because the requested change conforms to the criteria for actions eligible for categorical exclusion as specified in 10 CFR 51.22(c)(9). The requested change will have no impact on the environment. The proposed change does not involve a significant hazards consideration as discussed above. The proposed change does not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite. In addition, the proposed change does not involve a significant increase in the individual or cumulative occupational radiation exposure.
This limit is defined in EO-100/200-103.
ENCLOSURE C TO PLA-442S