ML20095A031
| ML20095A031 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/30/1995 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095A029 | List: |
| References | |
| NUDOCS 9512060283 | |
| Download: ML20095A031 (11) | |
Text
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. t UNITED STATES g
g NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20M6 0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.70 TO FACILITY OPERATING LICENSE NO. NPF-85 PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION. UNIT 2 DOCKET NO. 50-353 1
1.0 INTRODUCTION
By letter dated June 23, 1995, the Philadelphia Electric Company (the licensee) submitted a request for changes to the Limerick Generating Station (LGS), Unit 2, Technical Specifications (TSs). The requested changes would permit a one-time change affecting the Allowed Outage Time (A0T) for the Emergency Service Water (ESW) system, Residual Heat Removal Service Water I
(RHRSW) System, the Suppressior. Pool Cooling (SPC), Suppression Pool Spray (SPS), and Low Pressure Coolant Injection (LPCI) modes of the Residual Heat Removal System, and the Core Spray System to be extended from 3 and 7 days to 14 days during the Unit I refueling outage scheduled to begin in January 1996.
This proposed extended A0T ailows adequate time to install isolation valves and cross-ties on the ESW and RHRSW Systems to facilitate future inspections or. maintenance.
This TS request is similar to the LGS Unit 1, TS Change Request No. 94-04-1, issued by the NRC on January 27, 1995, as Amendment No. 86.
2.0 EVALUATION i
The licensee has proposed a one-time (i.e., temporary) change to extend the A0T for the ESW, RHRSW, SPC, SPS and LPCI modes of the RHR system, and Core Spray System.
The A0Ts are to be extended from 3 days to 14 days for ESW, RHRSW and SPC, and from 7 days to 14 days for LPCI, SPS, and core spray. The proposal is intended to allow adequate time to install isolation valves and cross-ties on the ESW and RHRSW systems, while continuing to operate Unit 2.
j The modifications will facilitate future inspections and maintenance by allowing isolation of one ESW loop from one unit while maintaining the operability of the remaining ESW loop to support operation of the other unit.
This evaluation assesses the adequacy of the proposed change and discusses the containment heat removal and depressurization functions of the RHR system (SPC, SPS), the acceptability of the temporary extension of the A0T, the risk impact based upon the incremental core damage probability due to the proposed activity, and the safety implications of extending the A0Ts for ESW and RHRSW.
9512060283 951130 PDR ADOCK 05000353 P
2.1 Containment Heat Removal /Depressurization Systems Design Bases 2.1.1 Suppression Pool Spray Mode of RHR Operation Design Basis: Following a Design Basis Accident (DBA), the RHR Suppression Pool _ Spray System (the Suppression Pool Spray mode of the RHR System) removes heat from the suppression chamber airspace in the suppression chamber above the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system resulting from a DBA or.a rapid depressurization of the reactor pressure vessel (RPV) through safety / relief.
valves. The heat addition to the suppression pool results in increased steam t
in the suppression chamber, which increases primary containment pressure. A small amount of the steam discharged into the drywell during a DBA can bypass the suppression pool and end up in the suppression chamber airspace without being condensed in the pool. This steam is known as suppression pool bypass leakage. The suppression pool airspace is sized to accommodate the noncondensible gases in the primary containment. Spraying cooled water into the suppression pool airspace provides a means to condense the bypass leakage steam.
The suppression pool spray function is provided by two of four independent RHR loops.
Each of these two (A and B) RHR suppression pool spray subsystems contains a pump and heat exchanger. The A and B pumps are backed-up by C and D pumps located in LPCI loops which lack heat exchangers.
Each pump has two 100% suction strainers. The two subsystems perform the suppression pool spray function by circulating water from the suppression pool through the RHR heat exchangers and returning the water to the suppression chamber via spray spargers and a return line. The spargers only accommodate a small portion (500 gpm) of the total RHR pump flow (10,000 gpm); most of the flow returns to e
the suppression pool through the suppression pool cooling return line. Thus, i
both SPC and SPS functions are performed when the SPS System is initiated. A minimum of one RHR SPS subsystem is required to mitigate potential bypass leakage paths and maintain the primary containment peak pressure below the design limits.
i Supporting cooling water systems: During the suppression pool spray mode of RHR operation, RHRSW, circulating through the tube side of the RHR heat exchanger exchanges heat with the suppression pool water and discharges this heat to the external heat sink. The RHR pump motor oil coolers and the RHR 4
compartments, are cooled using ESW. The loss of these supporting systems will result-in one train of suppression pool spray becoming inoperable.
(It is noted that raw RHRSW can be supplied directly to the suppression pool spray nozzles if necessary.)
Generic staff position on ACT:
From NUREG-1433, Rev.1, " Standard Technical Specifications, General Electric Plants, BWR/4," the generic staff position for a one-train A0T is:
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. With one RHR suppression pool spray subsystem inoperable, i
the inoperable subsystem must be restored to operable status within 7 days.
In this Condition, the remaining operable RHR suppression pool spray subsystem is adequate to perform the primary containment bypass leakage mitigation function. However, the overall reliability is reduced because a single failure in the operable subsystem could result in reduced primary containment bypass mitigation capability. The 7 day Completion Time was chosen in light of the redundant RHR suppression pool.
spray capabilities afforded by the operable subsystem and the low probability of a DBA occurring during this period.
The current Limerick TS requirement is consistent with the generic staff position.
2.1.2 Suppression Pool Cooling Mode of RHR Operation Design basis:
Following a DBA, the RHR SPC System removes heat from the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system blowdown.
In the long term, the pool continues to absorb residual heat generated by fuel in the reactor core. A means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits. This function is provided by the suppression pool cooling mode of operation of the RHR system. Water is drawn from the suppression pool by an RHR pump and circulated through an RHR heat exchanger to reduce its temoerature.
It is then returned to the suppression pool. This mode of operation is similar to the spray mode described above, and uses the same equipment, except that the suppression chamber spray spargers are not m,ed.
Supporting cooling water systems: The same support system requirements apply for SPC and for SPS mode.
Generic staff position on A0T:
From NUREG-1433, Rev.1, " Standard Technical Specifications, General Electric Plants, BWR/4," the generic staff position for a one-train A0T is:
With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to operable status within 7 days.
In this condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the operable subsystem could result in reduced primary containment cooling capability.
The 7 day Completion Time is acceptable in light of the redundant RHR suppression pool cooling capabilities afforded by the operable subsystem and the low probability of a DBA occurring during this period.
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i Although a permanent 7-day A0T would be acceptable under the terms of the new (NUREG-1433) generic position, the current Limerick TS specifies a 72-hour A0T.
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..l 2.2 Acceptability of Temporary Extension of A0Ts The proposed amendment would, for Unit 2, extend the A0T for TS 3.6.2.2, Suppression Pool Spray, from 7 days, 'to 14 days, one-time, for Loop A.
The A0T for TS 3.6.2.3 Suppression Pool Cooling Mode of RHR, would.be extended from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 14 days, one-time, for Loop A.
These two 14-day periods would coincide.
i During the proposed 14-day A0T period, both units will be subject to j
single-failure vulnerabilities that do not normally exist. The licensee has performed an analysis to identify single-active vulnerabilities. The components were found to be four motor-operated valves (MOVs), all located outside containment. The licensee then examined the maintenance history I
records for those MOVs. The maintenance records indicate that the " failure to l
operate on demand" history, for these valves was consistent with or less than the percentage failure rate (i.e.,1.18%) of the expected failure rate for MOVs (based on industry data). Based on this information, the staff concludes i
that the A0Ts may, on a one-time basis, as described in the application, be extended to 14 days which is 200% of the A0Ts specified in the STSs.
1 2.3 Conclusion The licensee has satisfactorily demonstrated that the SPC and spray functions of RHR may be granted a temporary, one-time 14-day A0T to permit modifications i
l that will improve the reliability of the ESW and RHRSW systems.
2.4 Risk Impact Discussion The licensee performed a Probabilistic Safety Assessment (PSA) for an ESW and RHRSW loop being inoperable for 14 days at power. The Core Damage Frequency 1
(CDF) increased by 3.14E-06/ reactor-yr (i.e., from 5.11E-06/ reactor-yr to 8.25E-06/ reactor yr). The corresponding incremental Core Damage Probability is approximately 1.2E-07, an acceptable level which is substantially less than the screening criterion level used to determine acceptability of Completion Times (i.e., A0Ts) in the new STSs.
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Additionally, the licensee is writing a special procedure to provide administrative control to maintain operability of specified components and implementation of appropriate compensatory measures necessary during the proposed 14-day A0T.
2.4.1 Conclusion The licensee's PSA analysis has established that the requested one-time A0T i
extensions to 14 days are acceptable to the staff, given that incperable i
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. i systems are returned to operable status as soon as possible and the new special procedure providing administrative control of component operability maintenance and compensatory measures during the activity is in place prior to the installation of isolation valves and cross-ties on the ESW and RHRSW.
Also, the plant modification will remove the need to remove an entire loop from service, reducing future plant risk.
2.5 Safety Implications of Extending A0T for ESW and RHRSW 4
2.5.1 RHR and Emergency Service Water RHRSW is a safety-related system, consisting of two independent sup' ply and return loops (A and B) common to both units.
Each loop services one RHR heat
- exchanger in each unit, and furnishes sufficient cooling for safe shutdown, cooling and accident mitigation in both units.
The RHR heat exchangers furnish decay heat removal (DHR) capability for normal shutdown cooling as well as primary containment suppression pool and drywell temperature and pressure control during accident mitigation.
Each loop has two pumps, each supplying 100 percent flow to one RHR heat exchanger.
The removal of either loop of RHRSW from service will impact the operability of the RHR heat exchangers in the affected loop.
I The ESW system is a safety-related system designed to supply cooling water to selected equipment during a Loss of Offsite Power or Loss-of-Coolant Accident (LOCA).
Safety-related equipment serviced by ESW includes: RHR pump motors, l
pump compartment unit coolers for RHR, Reactor Core Isolation Cooling (RCIC) system, High Pressure Coolant Injection (HPCI), and core spray, and standby diesel generator (SDG) heat exchangers. The system is common to Units 1 and 2, and is composed of two loops (A and B), with two 50 percent system capacity i
pumps per loop. The essential heat loads listed are automatically transferred to the ESW system under accident conditions, the ESW pumps start automatically on SDG operation and the pumps take suction on the spray pond.
Each SDG can 1
i be supplied with cooling water from 'A' or 'B' ESW. Normal alignment is such that loop 'A' furnishes cooling to the A and C diesel generators, and loop 'B' to the B and D diesel generators. Only two RHR pumps may be cooled by each ESW loop due to the RHR pumps' arrangement.
Each ESW return loop is connected to both ESW/RHRSW combined return headers to the spray pond, so that loss of a combined return header does not cause loss of an ESW loop.
The purpose of the proposed amendment is to support modifications to install jumpers, valves and crossties in the ESW and RHRSW systems during Unit 2 operation. The modifications will be performed during the upcoming Unit 1 refueling outage, during which the A loop of each system will be removed from service concurrently. The proposal involves TS 3.7.1.1, Action a.3 pertaining to an inoperable RHRSW subsystem, requiring one RHRSW subsystem restored with at least one RHRSW pump within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. An RHRSW subsystem consists of two RHRSW pumps and the flow path capable of cooling one RHR heat exchanger. Also involved is TS 3.7.1.2, Action a.3 pertaining to an inoperable ESW system loop, requiring the inoperable loop to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, The amendment would temporarily change the A0Ts to 14 days. TS 3.5.1 for
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. i Emergency Core Cooling System (ECCS) during plant operation is amended with a note linking the inoperability of LPCI
'B' to the shutdown cooling mode (SDC) lineup when reactor pressure is below the RHR SDC permissive set point. The A0T for LPCI in TS 3.5.1 is also extended to 14 days.
The following discussion considers operability implications of related ECCS and DHR equipment, as well as safety implications of SDG inoperability.
2.5.2 SDG Impact There are eight SDGs at LGS, four assigned to each unit. They are designated 5
by.two digits, the first indicating the unit, the second denoting the 1
generator. Without the
'A' loop of ESW, SDGs D11, D13, and D23 will be inoperable.
SDG D21 will be maintained operable since its heat exchanger will i
be aligned to 'B' ESW.
SDGs D12, D14, D22, and D24 remain operable, with 1
their normal source of ESW cooling available. TS 3.8.1.1 requires a minimum of four operable separate and independent diesel generators for the operating j
unit (Unit 2), and TS 3.8.1.2 requires a minimum of two SDGs operable for the shutdown of Unit 1.
The TS requirement for Unit 1 is satisfied with SDGs D12 i
i and D14 operable. Unit 2 will have one inoperable SDG, thus Action (a) of TS 3.8.1.1, with an associated 30-day A0T will be in effect during the extended-A0T period of 14 days.
Since the inoperability is due to preplanned i
maintenance, the licensee may demonstrate operability of the remaining Unit 2 j
SDGs (D21, D22, and D24) by performing surveillance requirement 4.8.1.1.la I
(breaker alignment and power availability checks).
This conclusion is based on the intent of the Action Statement as described in the LGS TS Bases and in the BWR4 Standard Technical Specification Bases (NUREG-1433, Rev. 1). Upon failure of a SDG, the concern of a possible common mode failure is addressed l
by testing of the remaining operable SDGs (per surveillance requirement L
4.8.1.1.2.a.4).
However, since the planned inoperability in this case does j
not introduce the possibility of common mode failures of the remaining Unit 2 l
SDGs, the staff considers the SDG startup tests unnecessary.
SDG inoperability is also addressed in the TS sections for RHRSW and ESW in i
terms of pump / diesel generator pairs. During the extended A0T, two RHRSW pump /SDG and ESW pump /SDG pairs will be inoperable for Unit 2, thus TS 3.7.1.1 i
(RHRSW) Action a.5 will be in effect, imposing a 30-day A0T. TS 3.7.1.2 ESW requirements are satisfied with two operable ESW pump /SDG pairs.
2.5.3 ECCS Impact ESW supports ECCS operation by providing cooling directly to some equipment, i
'to ECCS equipment room coolers, and to diesel generator heat exchangers. One loop of ESW is sufficient to furnish cooling to half of the ECCS equipment.
Per TS 3.7.1.2 Action a.3, which requires that all equipment aligned to the inoperable cooling loop be declared inoperable, removal of the 'A' ESW loop from service will result in the inoperability of loop
'A' core spray, and LPCI
'A' and 'C' loops.
If necessary, even without ESW cooling, the affected LPCI and core spray pumps could be operated on a short-term basis.
If equipment room temperatures became excessive (indications are available to control room i
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i personnel), operators could then follow procedures to reduce room temperatures.
Several hours are available to take actions to preclude equipment overheating.
Without the ESW 'A' loop cooling, the following operable ECCS equipment is available to respond to a LOCA: HPCI, core spray loop
'B', LPCI loops 'B' and
'D', and the Automatic Depressurization System (ADS). A modification implemented for HPCI room cooling (LGS Modification P-00212) will allow HPCI to remain operable throughout the extended A0T. The licensee has stated that i
the modification permanently eliminates the need for ESW as a HPCI support system.
Although not specifically included in the ESW system design basis, i
the cooling furnished by a single ESW loop is sufficient to safely shut down both units at all times except for the first 10 minutes of a postulated DBA
. LOCA. During the initial stage, three RHR pumps may be required to flood the LOCA unit using LPCI mode.
The combination of remaining ECCS equipment is not sufficient to accommodate the single failure evaluations listed in Updated Final Safety Analysis Report (UFSAR) Table 6.3-3.
Further, for an ECCS pipe break LOCA, the low pressure ECCS would not be single failure proof since a 'D' LPCI injection line break' and a failure of Division 2 AC power would incapacitate core spray 'B' and LPCI 'B' and 'D'.
However, while the plant is operating under a Limiting Condition for Operation (LCO), the minimum ECCS requirements are based on analyses presented in NED0-24708A, which is incorporated into the LGS UFSAR by reference. The generic analysis of DBA suction line breaks shows that one low pressure ECCS along with ADS is adequate to reflood the vessel and maintain core cooling and preclude fuel damage. The UFSAR discusses the outcome of LOCA studies documented in NED0-30936P which were done to establish success criteria for ECCS injection functions. Table 3-7 in NED0-30936P shows the minimum systems required to successfully respond to large, intermediate, and small LOCAs for BWR4 plants. The most restrictive combination shown is a i
single means of low pressure injection (LPCI or core spray) and ADS.
I The report stipulates (repeated in Section 6.3.1.1.2 of the LGS UFSAR) that an alternate cooling path may be needed for long-term LOCA cooling. This is i
because LPCI injects directly into the core shroud, which would maintain a substantial subcooling margin in the water around the core, thereby quenching 4
any steam cooling effects. The availability of core spray. satisfies the need for another means of long-term cooling.
In the event core spray is lost, F
i sufficient time exists for plant operators to take appropriate measures to establish other cooling means such as HPCI, ADS, SDC loop B, or the reactor i
water cleanup system.
Based on the generic analyses described above, the extension of the A0Ts for the systems affecting ECCS capability are acceptable.
Although the injection capability of the ECCS will not be affected when the
'A' RHRSW loop is inoperable, the 'A' RHR heat exchanger is declared inoperable.
The operability of RHR modes of operation dependent upon heat i
l exchanger operability will be affected and is discussed in the following section. The SPC and SPS mode of RHR were considered previously.
2.5.4 DHR Impact Two loops of shutdown cooling are required to be operable in accordance with TS Sections 3.4.9.1 (for hot shutdown) and TS 3.4.9.2 and 3.9.11.2 (for cold shutdown and refueling). Otherwise, an alternate method of decay heat removal is required.
While Unit 1 is shutdown and Unit 2 is operating, the Unit 1 SDC TS is satisfied with the operability of B and D loops of RHR.
If Unit 2 is required to shutdown, the 'B' and 'D' RHR pumps would remain in their LPCI mode alignment while depressurizing the plant to the RHR cut-in permissive set point. Once the set point pressure is reached, TS 3.4.9.1 Action a must be entered since only the 'B' SDC loop will be operable since it is capable of being aligned from the control room under normal shutdown conditions. The operability of alternate DHR methods would then be demonstrated in order to meet TS 3.4.9.1 requirements, following approved plant procedures, and would not be implemented unless SDC 'B' was rendered inoperable.
With the 'B' loop of RHR aligned to SDC, 'B' LPCI is then inoperable since the suppression pool suction isolation valve is shut. Thus, during the A0T, if 4
Unit 2 is to be shutdown, three LPCI loops and a core spray loop would be inoperable, a condition not addressed in TS 3.5.1 Actions a.1 and b.5.
The licensee is proposing a change to TS 3.5.1 which adds a note addressing this situation for the extended A0T. The note links the inoperability of LPCI 'B' to the SDC lineup when reactor pressure is below the RHR SDC permissive set point, permitting this situation during the extended A0T. Under the conditions imposed by a Unit 2 shutdown, sufficient ECCS (i.e., LPCI 'D' and core spray 'B') are available, since an active single failure is not postulated during the A0T and only one low pressure ECCS is necessary per i
NED0-24708A and NED0-30936P, as discussed earlier.
The TS modification to accommodate the inoperability of LPCI during Unit 2 shutdown is considered acceptable since the plant is not expected to remain in hot shutdown for an extended duration and a ! ngle low pressure system will be available for LOCA response as discussed prev. usly in this evaluation.
Once Unit 2 is in cold shutdown, both the 'B' and 'D' RHR loops will be considered operable for SDC according to TS 3.4.9.2, and TS 3.5.2 ECCS requirements for plant :hutdown will be met by having a core spray and a LPCI loop available.
l 2.5.5 Impact on Other Systems The application included evaluations of the safety impact of reduced RHRSW and ESW operability on a number of other systems, including non-safety related
systems, due to reduced cooling or loss of associated SDGs. Among the safety-related systems evaluated were:
- RCIC, Fuel Pool Cooling, Main Control Room Chillers, Control Room Emergency Fresh Air System (CREFAS),
Standby Gas Treatment System (SGTS),
Reactor Enclosure Recirculation System, Standby Liquid Control, Hydrogen /0xygen Analyzers, Containment Hydrogen Recombiners, Main Steam Isolation Valve Leakage Control, Drywell Mixing System, Spray Pond Pump House Heating, Ventilation, and Air Conditioning, Safeguard Piping Fill System, Fire Protection, Batteries and Battery Chargers, Remote Shutdown System.
Except for the systems or functions asterisked, the licensee determined that
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if the system operability was affected (in many cases the systems remained fully operable), that TS requirements were satisfied and that no related Action Statements were entered.
j The effect on fire protection capability does not place TS restrictions on the plant, but as discussed in the Compensatory Measures Section of this evaluation the licensee is instituting measures to assure that adequate fire protection is available during the extended A0T period. The inoperability e
intrcduced by loss of ESW cooling capability to the other asterisked systems places the plant in 30-day A0Ts, which are bounded by the proposed extended A0T period of 14 days. The licensee also evaluated the single failure implications for equipment left operable during the A0T in the case of a combined LOCA/ LOOP. Such an evaluation was a prudent measure, although single failures are not required to be considered during an A0T. Thus, operation under the LC0 Action Statements considered here do not pose an adverse safety impact during this one-time A0T extension. The licensee evaluation of the risk impact determined through 1 PRA reinforces that conclusion.
2.5.6 Compensatory Measures The licensee intends to implement a number of compensatory measures during the extended A0T period to offset the reduction in safety margin resulting from the out-of-service systems. A special procedure, will be enacted to control plant activity for the duration of the extended A0T.
Significant among the steps to be taken are:
a.
In order to maintain SDG D21 operability, ESW cooling water for the SDG D21 will be realigned so that 'B' ESW loop is the cooling source.
i I b.
The ESW return valve to the 'B' RHRSW return header will be verified blocked open, and the ESW return valve to the 'A' RHRSW return header will be blocked closed, prior to entering the A0T. This is done to limit single failure potential during the extended A0T.
The licensee will follow established plant procedures (the LGS Clearance and Tagging Manual) to carry out the specific valve blocking guidance in the special procedure.
c.
Operability of specifically listed equipment necessary to support power operation of Unit 2 will be verified prior to implementing the extended A0T.
d.
Unit 2 suppression pool temperature is to be maintained low, (the licensee stated objective is below 80*F) for the duration of the extended A0T.
e.
Work on the 'A' RHRSW loop will not commence until the Unit I decay heat generation rate is within the capability afforded by the remaining equipment ('B' RHR heat exchanger) is available, as specified by plant l
procedure.
f.
The Unit I reactor cavity will not be drained until the
'A' RHRSW loop is j
returned to service or an alternate DHR method is available.
Procedures to establish an alternate heat removal capability will be incorporated into the special procedure.
g.
The Unit 2 SDGs D22 and D24 as well as the Unit 1 SDG D12 must be maintained operable to maintain the full operability of the 'B' RHRSW and
'B' ESW loops.
h.
The Unit 1 SDGs D12 and D14 must be maintained operable to maintain the full operability of the CREFAS and SGTS 'B' loops.
i.
The Unit 2 SDG D21 and the Unit 2 Division I DC battery system must be maintained operable to support the use of Division I ADS trip system subsystem 'A' for ECCS and DHR.
J.
The Unit 1 SDG D14 must be maintained operable to support the fire protection non-safeguard loads.
These compensatory measures, along with the licensee's commitment to limit the performance of surveillances during the extended A0T, support the contention i
that the safety impacts of the reductions in RHRSW and ESW capabilities are being minimized.
The staff finds these steps to be adequate for the extended A0T duration.
2.5.7 Conclusion The staff has evaluated the safety implications of the proposed one-time extension of the A0Ts for ESW, RHRSW, the LPCI mode of RHR, and core spray during modification of the ESW and RHRSW systems. The conclusions reached on containment cooling and depressurization functions, and on the licensee's risk i
assessment were considered for this review.
The licensee's appraisal of the operability impact of the systems listed in Section 2.5 of this evaluation
o
. i appears sound.
Based on the evaluation of the safety-related system capabilities maintained, compensatory measures instituted using a special procedure, and the licensee's commitment to limit surveillances, the staff finds the proposed TS changes acceptable on the limited one-time basis required during the Unit I sixth refueling outage.
3.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIR 6 The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase.in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no puolic comment on such finding (60 FR 39448). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR Sl.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, j
and (3) the issuance of the amendment will not be inimical to the common i
defense and security or to the health and safety of the public.
Principal Contributors:
J. Donoghue W. Long M. Wohl F. Rinaldi S. Dembek Date: November 30, 1995 1
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