Text

Safety Evaluation Supporting Amends 174 & 178 to Licenses DPR-24 & DPR-27,respectively
	ML20217N374
Person / Time
Site:	Point Beach
Issue date:	07/09/1997
From:	; NRC (Affiliation Not Assigned)
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ML20217N373	List: ML20217N367; ML20217N374;
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UNITED STATES y*

p NUCLEAR REGULATCRY CSMMISSION WASHINGTON, D.C. Sett6 4eM

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SArETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REL ATED TO' AMENDMENT N05.174 AND 178 TO FACILITY OPERATING LICENSE NOS. OPR 24 AND DPR 27 i

WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT. UNIT NOS. 1 AND 2 DOCKET NOS. 50 266 AND 50 301 1.0-INTRODUCTION By letter dated September 30,1996 (TSCR 192), as supplemented on November 26 and December 12, 1996, February 13, March 5. A97 0350 and 97 0351)pril 2, April 16. May 9. Ju June 13 (two letters, NPL

, and June 25.-1997, the Wisconsin Electric Power Company (the licensee) requested amendments to the Technical 5)ecifications (TS) appended to Facility Operating License Nos.

DPR 24 and )PR 27 for the Point Beach Nuclear Plant (PBNP), Units 1 and 2.

The proposed amendments would change TS 15.3.3, " Emergency Core Cooling System Auxiliary Cooling Systems, Air Recirculation Fan Coolers, and Containment Spray." to incorporate allowed outage times (A0Ts) similar to those contained in NUREG 1431 Revision 1. " Westinghouse Owner's Group Improved Standard Technical Specifications," and modify the operability requirements for the service water (SW) and component cooling water (CCW) systems. These pr posed changes are intended to ensure that TS requirements will reflect the owest functional capability or erformance levels of equ pment required for safe operation of the facil ty," as defined in Section 50, 6(c)(2), "Limitin Conditions for 0)eration (LCO)." Additional changes were proposed to TS 1.3.7, " Auxiliary Electrical Systems," reflecting the proposedSWsystem(SWS)operabilityrequirements: to TS 15.3.12. " Control Room Emergency Filtration, revising charcoal filtration efficie cies and including an appropriate testing standard in the TS basis: and to TS 15.5.2.C.2 lowering the design heat removal capacity of tint los r cont +ont ventilation fans and air cooler units.

The June 13 and June 25, 1997, letters provided clarifying information w* thin the scope of the application and did not affect the staff s previous no significant hazards considerations determinations.

The March 5,1997, submittal identified the following licensee comitment regarding actions to be taken for the control room ventilation system.

Wisconsin Electric will achieve compliance with the dose limits associated with 10 CFR 50 Appendix A. GDC 19, which states:

A control room shall be provided from which actions can be taken to operate [the]

nuclear powe? unit safely under normal conditions and to maintain it in Qb0 W

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2 a safe condition under accident conditions, including loss of coolant accidents. Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation ex)osures in excess of 5 rem whole body, or its equivalent to 'any part of tie body, for the duration of the accident.

Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2)-with a l

potential capability for subsequent cold shutdown of the reactor through j

the use of suitable procedures.

l-

-The June 25, 1997, submittal also identified the following licensee commitment-regarding the operation of the SWS:

Wisconsin Electric will operate Point Beach Nuclear Plant in accordance with its service water system analyses and approved procedures.

Specifically, each unit will utilize only one CCW heat exchanger until such time that analyses are completed and the service water system reconfigured as necessary to allcw operation of one or both units with two heat exchangers-in service.

If two CCW heat exchangers are required in one or both units for maintaking acceptable component cooling water temperature prior to completion of necessary analyses to allow operation in the required configuration, the service water system will be considered in an unanalyzed condition, declared inoperable and action taken as specified by Technical Sper.ification 15.3.0 B except for short periods of time as necessary to effect procedurally controlled changes in system lineups and unit operating conditions.

This condition was needed to support normal operation with two heat exchangers in service when SWS temperature (at the inlet to the CCW heat exchangers) reaches about 71 degrees Fahrenheit ('F) based on operating history.

The licensee has proposed the following specific changes to the plant TS.

a.

TS 15.3.3.B.2 currently requires that during power operations the requirements of 15.3.3.B 1 may be modified to allow any of the following

. components to be inoperable at any one time:

(a) one accident fan cooler may be out of service provided that cooler is returned to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and the other accident fan coolers shall-be operable before initiating maintenance on the inoperable fan cooler, (b) one containment spray pump may be out of service provided the pump is restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and the remaining containment spray pump shall be operable before initiating maintenance on the inoperable pump, and (c) any valve required for the functioning of the system during accident conditions may be inoperable provided repairs are completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

4

3-The licensee proposes the following changes to TS 15.3.3.B.2:

Instead of allowing "any of the followin inoperable" the revised TS would allow "g components to be any one of the following limitations (a, b, or c) to be in effect at any one time."

Limitation a. would be changed to allow one or two accident fan coolers to be inoperable with an A0T of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Limitation b. would be changed to delete the abrase that specifies "before initiating maintenance on any ino> era)1e pump," and add a sentence to Limitation (a) that reads, "T1e remaining accident fan coolers shall be operable."

Limitation c. would be revised to include an exce) tion that reads "If a spray pump is removed from service per b, a>ove, valves associated with that train may"be removed from service for the period specified for the pump, b.

TS 15.3.3.C. " Component Cooling System." currently has separate requirements for single unit and two unit operation. For single unit operation the limiting condition for operation (LCO) specifies that one reactor shall not be made critical unless the following conditions are met:

a.

The two CCW pumps assigned to that unit are operable, b.

Both the CCW heat exchanger associated with the unit together with one of the shared saare heat exchangers are operable or the two shared spare leat exchangers are

operable, c.

All valves, interlocks and piping associated with the above components, and required for the functioning of the system during accident conditions, are operable.

For two unit operation the LCO specifies that both reactors shall not be made critical unless the following conditions are met:

a.

Three component cooling pumps are operable, b.

Three component cooling heat exchangers are operable.

c.

All valves, interlocks and piping required for the functioning of the system during accident conditions and associated with the above components are operable..

For single unit-operation, the current specification allows one of the assigned CCW pumps to be inoperable for u) to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months and allows two heat exchangers which may be aligned to tie operating unit to be out of service for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

For two unit operation the current

4 specification allows one of the three assigned CCW ) umps to be out of service for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and allows two heat exc1 angers to be out of service for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

'The proposed changes to TS 15.3.3.C would cor6bine the separate single and two unit LCOs under this s)ecification into one LCO.

The single LC0 wpuld specify that a reactor s1all not be made critical unless the

.ullowing conditions are met:

a.

The two component cooling pumps assigned to that unit are operable.

b.

Either the com)onent cooling heat exchanger associated with the unit togetler with one of the shared spare heat exchangers are operable or the two shared spare heat exchangers are operable for single unit operation.

Three component cooling heat exchangers are operable for two unit operation.

c.

All valves, interlocks and piping associated with the above components, and required for the functioning of the system during accident condition, are operable.

The proposed changes would allow one of the required CCW pumps to be out of service for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and would allow one of the required heat exchangers to be out of service for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , c.

TS 15.3.3.D.1 currently requires four SW pumps, two from each train, and all necessary valves, interlocks and piping required for the functioning of the SWS during accident conditions be operable during power operations.

The licensee proposes to change the number of SW pumps required to be operable from four to six and dehte the requirement that states. "two from each train."

d.

TS 15.3,3.D.2 currently allows that during power operation, the requirements of 15.3.3.0.1 may be modified to allow one of the following components to be inoperable at any one time:

(a) one of the four required SW pumps may be out of service provided a pump is restored to operable status within 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s: (b) one of the two loop headers may be out of service for a period of up to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s: (c) a valve or other passive component may be out of service provided repairs can be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

If any of the requirements are not met the LC0 requires both reactors to be placed in the hot shutdown condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The licensee proposes the following changes to TS 15.3.3,0.2:

Replace the statement to allow one of the following components to be inoperable at any one time with a statement that would allow any of the proposed conditions to be in effect coincidentally for the A0T identified in the condition.

Each of the conditions i

O

.5-includes provisions to ensure adequate flow capability remains available for coincidental entry into more than one condition.

Change the action statement to require the affect,ed reactor (s) to be shut down rather than both units.

Change TS 15.3.3.D.2.a to state, "One of the six required service water pumps may be out of service provided a pump is restored to operable status within 7 days. A second service water pump may be out of service provided a pump is restored to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . A third service water aump may be out of service l

)rovided two pumps are restored to operaale status within 72 1ours."

Change TS 15.3.3.D.2.b to state that the SW ring header continuous flowpath may be out of service for up to 7 days and include the l

provision that if less than four SW pumps are operable. SWS flow shall be evaluated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of entry into this LC0 and the LCOs for the affected equipment would be entered.

Change TS 15.3.3.D.2.c to apply to any or all automatic isolation valves required during accident conditions rather than "a valve or other passive component" and increase the A0T from 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least four SW pumps are operable.

Additionally, this condition can be exited provided the affected line is isolated with a seismically qualified isolation valve or if the valve is restored to operable status.

Insert new specification 15.3.3.D.2.d. to allow the containment fan cooler outlet motor operated valves to be o)en for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least five SW pumps are operaale.

e.

Change TS 15.3.3 Basis to include supporting information for each of the previously described changes.

f.

TS 15.3.7. A.1 currently specifies in part, the following LCOs:

"i.

For both units to be made critical. the normal power supply and a standby emergency power supply to all the 4160/480 Volt safeguards buses shall be operable and the buses are energized from their normal supply.

"j.

For Unit I to be made critical, the normal power supply and c standby emergency power supply to the 4160/480 Volt safeguards buses Unit 1 A05/B03. Unit 1 A06/B04, and Unit 2 A06/B04 shall be operable and the buses are energized from their normal supply.

"k.

For Unit 2 to be made critical. the normal power supply and a standby emergency power supply to the 4160/480 Volt e'

safeguards buses Unit 2 A05/B03. Unit 2 A06/B04, and Unit 1

6-A05/B03 shall be operable and the buses are energized from their normal supply."

The licensee pro)oses to change TS 15.3.7.A.1.1 to require that the LC0 a) ply to one 'or soth units rather than just both units.

As a result of t1e change to LCO 1.. the licensee is proposing to delete both LCOs j.

and k.. which are no longer necessary, g.

TS 15.3.7.B.1.9 currently allows the normal aower supply or standby emergency power supply to Unit 1 A06/B04 or Jnit 2 A05/B03 or both to be out of service for up to 7 days (provided the redundant engineered safety features (ESF) are operable and the required redundant standby

)

emergency power supplies are started within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months l

thereafter).

It also specifies that if the normal power supply is out l

of service, an operable emeegency diesel generator is supplying the t

affected 4160/480 Volt busw.

After 7 days, the affected unit or both units will be placed in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and I

cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The proposed change requires both units, rather than the affected unit or both units, to be

) laced in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown wit 11n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

h.

TS 15.3.12. " Control Room Emergency Filtration." currently requires laboratory charcoal adsorbent tests, conducted in accordance with TS 15.4.11 shall show a minimum of 90% removal of methyl iodide.

The Basis for TS 15.4.11 currently does not specify a standard to perform the charcoal filter testing.

The proposed change revises the acceptance criteria for charcoal adsorbent tests to a minimum of 99% removal of methyl iodide.

The licensee pro)oses to include a statement in the Basis for TS 15.4.11 to ensure the clarcoal adsorbent laboratory sample analysis is performed in accordance with ASTM [American Society for Testing and Materials]

D3803 89. " Standard Test Method for Nuclear Grade Activated Carbon."

1.

TS 15.5. " Design." Section 15.5.2.C.. " Containment Systems." Item 2..

states. "The containment vessel has an internal air recirculation system which consists of four ventilation fans and air coolers capable of a total heat removal capability of 55.6000 Btu /sec under conditions following a loss of-coolant accident."

The proposed TS change also results in a reduction in the total heat removal capability of the four containment fan coolers (CFCs) from 55.600 Btu /second to 41.700 Btu /second.

2.0 BACKGROUND

Licensee Event Report (LER) 96 004-00, dated August 30, 1996, and its supplement. LER 96 004 01, dated September 11, 1996, describe the discovery of conditions that placed the plant outside of its design basis.

The conditions

7-involved the ability of the SWS to provide sufficient flow to components relied upon to mitigate the consequences of a design basis accident (DBA).

In LER 96 004 and its supplement, the licensee documented its determination that three SWS pumps were required (in lieu of two SWS pumps as determined by previous analyses) fo'r adequate heat removal capability following 'a design basis loss-of coolant accident (LOCA).

Therefore, in order to meet single failure requirements, six SWS pumps were required to be operable (rather than four as currently required in the existing specifications) to adequately mitigate the consequences of a DBA.

On June 24, 1996. Westinghouse issued a Nuclear Safety Advisory Letter (NSAL) 96 03. " Containment Fan Cooler Operation During a Design Basis Accident." NSAL 96 03 described the potential for steam flashing in the CFCs during a DBA.

The licensee determined + hat the issues identified in NSAL 96-

30. pplied to Point Beach, and as a result issued LER 03 a 96-005 00, dated August 1996.

LER 96 005 describes the potential for SW flashing within the CFCs during a design basis LOCA with a concurrent loss of offsite power (LOOP), and the potential for water hammer in the SWS following the normal sequencing of l

safety related equipment onto the emergency buses.

The issues identified in LER 96 004 and its supplement only addressed steady state conditions (full SWS flow and full CFC heat removal function) l that were considered in the original design basis.

The issues identified in LER 96 005 and described in NSA'. 96 03 were not considered in the original design basis and relate to the transient conditions (about the first minute) that occur following a LOCA/ LOOP and are associated with the tripping and restart of the CFC fans and SWS pumps.

During this transient condition boiling occurs in the CFCs resulting in the potential for a delay in CFC heat removal function in addition to the potential for an SWS water hammer after restart of the SWS pumps. To ensure no boiling (two phase flow) occurs during the longer term steady state conditions, the licensee determined that the SWS needed to be modified by throttling the SWS discharge valves from the CFCs.

This throttling of the discharge valves raises the pressure in the CFC cooling coils. However, the throttling of the discharge valves reduces the ability of the CFCs to remove heat from containment.

In the licensee's revised containment analysis. CFC heat removal capability was conservatively assumed to be reduced by approximately 25% (from 55.600 to 41.700 Btu /second).

By letters dated September 9 and September 30, 1996, the licensee also submitted detailed SWS operability evaluations addressing the post LOCA transient and steady state conditions, respectively.

The operability evaluation covering tha transient conditions was intended to show that the resulting water hammer would not result in loss of SWS integrity based on interim operability criteria.

The operability evaluation covering the steady state conditions was intended to show adequate long term (after 1 minute) SWS flow to support the revised containment analysis.

As a result of the concerns identified in NSAL 96 03 the staff issued Information Notice 96-45 " Potential Common Mode Post-Accident Failure of Containment Coolers." dated August 12. 1996, followed by Generic Letter (GL) 96-06. " Assurance of Equipment Operability and Containment-Integrity During j

8

' Design Basis Accident Conditions," dated September 30, 1996.

In GL 96 06, the staff requested licensees to determine if containment air cooler cooling water systems are susceptible to either water hammer or two phase flow conditions during postulated accident conditions.

If the systems were found to be susceptible to the conditions discussed in the GL, the licensees were to assess the operability of the affected systems and take corrective actions as appropriate. The licensees were specifically requested to submit (within 120 days) a summary report stating the conclusions that were reached relf ae to susceptibility for water hammer and two phase flow in the containment air cooler cooling water system, the buis for continued operability of affected systems and components, and corrective actions that were implemented or planned to be implemented, The licensee's response to GL 96-06 referenced the operability evaluations submitted on September 9 and September 30, 1996.

The staff performed a areliminary review of those operability evaluations as part of its-review of t1e proposed TS changes that are the subject of this safety evaluation.

The licensee's revised post LOCA heat removal capability analysis included information )rovided in LER 90-011-00, dated September 27, 1990, which identified tlat, under certsen conditions, the containment spray (CS) pumps could not take suction (due to lack of adequate net positive suction head

[NPSH]) from the residual heat removal (RHR) pump discharge during the recirculat'.on mode of operation (see evaluation of changes to TS 15.3.3.B.2 in Section 3.1 of this evaluation).

In the most recent re-analysis the licensee determined that to meet design requirements for containment heat removal after a DBA, the RHR heat exchanger, cooled by the CCW system, was required to I

operate in the recirculation mode when the CS ) umps could no longer take suction from the RHR Sumps (abnut I hour post.0CA). Therefore, the licensee proposed changes to t1e CCW system TS to ensure that after a postulated large-break LOCA, with a LOOP. and the worst single failure (loss of one 4160/480 Volt bus), the minimum number of CCW pumps remained o>erable without the need to rely on cross-connections with the CCW system of tie non accident unit.

The changes to the heat removal capability (25% capacity reduction in CFC heat removal capability) and the revised CS pump operation resulted in changes to the pressure and temperature profiles included in the Final Safety Analysis Report (FSAR). The changes in the containment pressure profile and the shorter assumed duration of CS pump operation required a reevaluation of radiation doses that could be received by an individual in the low population zone (LPZ), exclusion area boundary (EAB), and control room.

The= licensee also evaluated the impact of the changes to the containment pressure and tempercture profiles on the environmental qualification of

. equipment inside of containment in addition to the effects on the operction of the post-accident sampling system (PASS).

i The following. evaluation provides the results of the staff's review for each of the proposed TS changes.

It also includes the results of the staff's review of the licensee's containment heat removal analysis including the revised containment pressure and temperature profiles (change to

__=

9 TS 15.3.3.D.2), the licensee's SWS operability determine N m (change to TS 15.3.3.D.1), and the results of the staff's review of the w 'ssments provided for the environmental qualification of equipment (Section 2.2) and the effect on the PASS operation (Section 3.3). The staff's evaluation of the changes to radiological consequences resulting from the revised hnalysis are provided in Section 3.4 of the evaluation.

The staff's evaluation of the licensee's operability determinations (September 9 and September 30, 1996, submittals) is preliminary in nature.

The staff will perform a detailed evaluation of the licensee's o)er*5111ty submittals as.part of its review of the licensee's response to G. WOG. The results of this preliminary evaluation are included in the evcluation of the licensee's proposed change to TS IS.3.3.D.1.

3.0 EVALUATION The CFCs of both units are supplied cooling water by the shared SWS during normal operating and accident conditions. In the event of an accident in one unit, the SWS flow to that unit's CFCs is automatically increased by the opening of redundant valves on the discharge side of the CFCs. The SWS flow arrangement to the non-accident unit's CFCs would not change and the flow would be dependent on the number of running SWS pumps. The SWS includes six shared motor-driven SWS pumps (three pumps powered from Train A [ emergency diesel generator G-01] and three pumps powered from Train B [ emergency diesel generator G-02]) which are connected in parallel to a common discharge header.

This common discharge header sup) lies flow to two main SV" headers (north and south).

Both the north and sout, headers feed a common sup)1y header (west) which can be sup)1ied full flow by either the north or souti header alone if one of the main leaders is isolated. All three headers (north, south, and west) supply cooling water flow to a number of shared safety-related components. The south header supplies flow to a number of Unit 1 only components (including two of the four Unit 1 CFCs) while the north header serves a number of Unit 2 only components (including two of the four Unit 2 CFCs). The common west header supplies cooling water flow to the remaining two CFCs of both units in addition to a number of shared components.

Therefore, with either of the three headers totally isolated. operation of one or both units can be affected.

Isolation of the west header would affect both

-units while isolation of either the north or south header would affect the respective unit whose CFCs-can only be supplied by that header.

The SWS headers also supply cooling water flow to four CCW heat exchangers, two of which are dedicated (heat remo M from only one unit's CCW system) to specific units and two of which are shared spare heat exchangers that can be lined up to the CCW system of either unit. All four heat exchangers (normally only two heat exchangers (one for each unit) are in service with SWS flow lined up) are supplied cooling water from both the north and south SWS headers.

Following a DBA in one unit, only one CCW heat exchanger and one CCW pump are required to remove design-basis heat loads in the accident unit and only one heat exchanger and one CCW pump are required to maintain safe shutdown conditions in the non-accident unit.

Each unit has its own closed loop CCW system with two pumps and normally the two units' CCW systems operate

independently of each other. Cross-connect capability exists between-the two

-units such that all four CCW pumps can supply the CCW heat loads of either unit. Under the current CCW specifications, this cross-connect capability is relied upon to meet single active failure criteria because only three CCW pumps are requi' red for two unit operation.

If a single active failure occurred that affected the operation of the CCW loop with only one pump, operator action would be required to cross-connect to the other unit s CCW loop. With the proposed changes to the CCW specifications, the cross-connections between the two units are no longer relied upon because each unit will be required to have two operable CCW pumps.

Each unit's CCW loop, in addition to other safety-related heat-loads, supplies cooling water to two RHR heat exchangers. The revised containment cooling analysis relies on the RHR heat exchangers for cooling during the recirculation phase of the design-basis LOCA.

3.1 Technical Soecification Chances a.

'TS 15.3.3.B.2 - The current TS requires that, for each unit, all four containment fan coolers be o)erable during power operation, but permits one fan cooler to be inoperaale for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This LC0 would be changed to specify that all fnur coolers be operable during power operation but that one or two fan coolers may be inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Also, a requirement that the other fan coolers be operable before initiating maintenance on any inoperable fan cooler would be changed to delete the phrase that specifies "before initiating maintenance on any inoperable fan cooler." This phrase would be removed because it is unnecessary, A similar statement applicable to spray pumps would also be deleted.

The proposed TS changes are consistent with the standard generic criterion that a safety system such as the containment cooling system shall be redundant (e.g., not subject to single failures), but short periods of time up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continued operation in a degraded-(vulnerable to system or su) port system single failure) are permitted in order to restore an inopera)le subsystem.

Thestandardgenericcriteria for completion times in the event required equipment is inoperable are as follows:

7 days of continued operation is permitted while in a degraded condition if an additional single failure-(including loss of an AC power subdivision) could be tolerated.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continued operation is permitted if the operable containment cooling systems are sufficient.

Commence a shutdown if a safety function is completely lost.

Based on consistency with the second criterion, the requested 72-hour-completion time for restoration of one or two inoperable fan coolers is acceptable.

The licensee's application, including supplemental information, describes new containment pressure / temperature analyses that were performed to determine the adequacy of the containment cooling system with reduced SW flow to the fan coolers.

The containment pressure and temperature analyses'were performed by the nuclear steam supply system (NSSS) vendor using an updated C0C0 code model.

C0C0 is the NSSS vendor's approved code for containment post-accident long-term pressure and temperature response.

The significant differences between the new C0C0 model and the original (ca. 1968) licensing model are:

1.

Use of American Nuclear Society (ANS)-5.1-1979 decay heat values versus ANS 5.1-1971.

2.

Fan cooler capacity reduced to 75% of original assumption.

3.

Termination of spray pump operation at I hour.

4.

RHR heat exchanger provides 33 'F cooling of sump fluid. Original model assumed refueling water storage tank (RWST) water supply throughout the event.

The use of ANS-5.1-1979 for decay heat values is consistent with the "American National Standard - Pressure and Temperature Transient Analysis for Light Water Reactor Containments." American National Standards Institute (ANSI)/ANS-56.4-1988.

It is used with a 2-sigma uncertainty.

The staff's 30sition is that ANS-5.1-1979 decay heat model is acceptable for use for

)BNP.

The fan cooler heat transfer capacity reduction is bt. sed on SWS modifications analyzed by the licensee using a systen hydraulic model.

The heat transfer reduction is acceptable based on the reduction in SW flow to the fan coolers.

Termination of spray pump operation at 1-hour is a result of NPSH limitations previously reported in LER 90-011. As described in the LER, the RHR pump cannot supply sufficient flow at sufficient pressure to supply both the safety injection pumps and the CS pumps.

It is noted that the June 3,1997, supplement indicates that, for the limiting scenario, spray would be terminated at 69 minutes. Sixty-nine minutes of spray o>eration would provide additional containment cooling and is therefore )ounded by the 60-minute assumption.

The 33 'F cooling assumption for RHR heat exchanger results in a colder emergency core cooling system (ECCS) injection temperature and thus a reduced line break mass and energy flow into the containment.

In previousanalyses,alldecayheatenteredthecontainmentatmospherein the form of steam. The 33 F figure is based on the assumption that the containment sump temperature remains constant following recirculation

switchover. This is an acceptable modeling feature and is consistent with the NSSS vendor's containment analysis methodology.

4 The staff concludes that the new analyses indicate that (a) containment post-LOCA long term cooling fs adequate to prevent containment overpressure, and (b) the containment pressure response upon which containment leakage and associated radiological dose calculations are L

-based remains valid, b.

TS 15.3.3.B.2.c - The licensee proposes to add the following to TS 15.3.3.B.2.c. " Exception:

If a spray pump is removed from service per

b. above. valves associated with that train may be removed from service for the period specified for the pump " This statement was removed i

under Amendments 159 (DPR-24) and 163 (DPR-27). This statement is L

-needea because 15.3.3.B allows only one LC0 for containment cooling at L

any one time, which previously had the logical exception for the spray pumps and valves in the same flow train.

The staff finds the addition of the exception acceptable since it provides clarifying information, c.

TS 15.3.3.C - The proposed change to TS 15.3.3.C.1 would eliminate the one unit and two unit operability requirements for the CCW system and specify two operable CCW pumps aer unit, for a total of four CCW pumps for two unit operation rather tlan the current TS requirement of three CCW pumps for-two unit operation.

The proposed change to TS 15.3.3.C.2' would allow one of the two required CCW pumps 3er unit to be out of service provided the pump is restored to opera)le status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The current saecification allows one pump to be out of service for only 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . T1e increase in the number of operable CCW pumps is consistent with the new containment analysis which requires CCW flow to the RHR heat exchanger to provide containment cooling and 3ressure reduction following a design-basis LOCA with a concurrent

.00P and single active failure.

The proposed change requires an additional pump to be operable for operation of both units and eliminates the reliance on the cross-connections between the two units' CCW systems. The proposed change-is, therefore, more conservative than the existing specifications. The 72-hour A0T is based on the redundant capabilities afforded by the remaining operable CCW equipment and the low probability of a DBA during the 72-hour period. The 72-hour A0T is consistent with the A0Ts for other ESF equipment at Point Beach and is also consistent with NUREG-1431.

Based on the above, the staff nas determined that the proposed changes related to the number of CCW pumps and the ADTs for the CCW pumps are acceptable.

The 3roposed LCOs for the CCW heat exchangers are essentially identical to t1e current specifications and will require the CCW heat exchanger--

associated with the operating unit and one of the shared saare heat exchangers to be operable or the two shared spare heat exc1 angers to be operable for single unit operation. Three of the four CCW heat exchangers are required for two unit operation. The licensee also proposes to increase the A0T for one heat exchanger from 48 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

During normal plant o)erations one CCW heat exchanger and one CCW pump are in service at eac1 unit.

During elevated SWS temperatures (about

13 -

71 'F based on past history) two heat exchangers may be placed in service to maintain CCW temperature within operating limits. Also, during cooldown of a unit, two CCW heat exchangers are in service until some time after cold shutdown conditions are reached.

On June 11, 1997.

the licensee notified the staff that if'a DBA were to occur in an operating unit with two heat exchangers on either unit in service, the potential for a SWS flow diversion exists (too much flow going to the CCW heat exchangers) such that other SWS heat loads may not get adequate flow to meet design basis heat removal rates.

As a result of this

-recent determination, the licensee, in its June 25, 1997, submittal.

committed to operate with only one CCW heat exchanger lined up to each unit until adequate analyses have been completed as described in Section 1.0 of this safety evaluation.

The change to the CCW heat exchanger TS is consistent with the current specification, and since one heat exchanger is specific for Unit 1. one heat exchanger is s)ecific to Unit 2. and the remaining two heat exchangers can be slared between units, sufficient redundancy is provided to accept an increase in the A0T from 48 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The A0T of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is also consistent with the A0Ts for a CCW system heat exchanger as presented in NUREG-1431.

Based on the existing redundancy and the licensee's identified commitment on use of two heat exchangers when the ultimate heat sink temperature is high the staff finds the changes acceptable.

I d.

TS 15.3.3,0.1 - The licensee's proposed change would increase the number of SWS pumps required to be operable from four to six and would delete the requirement that states "two [ pumps] from each train " The total number of pumps available in the SWS is six; therefore, the proposed change is conservative because it requires all the pumps in the system to~be operable. The proposed deletion of "two from each train" is no longer required because all six pumps are required to be operable and that includes three sup) lied from each electric bus, The licensee proposed to eliminate tie )hrase "or maintained critical"'since the t.COs allow specific times when laving less than six operable SW pumps is acceptable. The requirement for six pumps is necessary to support the latest containment analysis. The proposed change is acceptable because it increases the number of required pumps and is necessary to be consistent with the design basis of the plant and the revised containment analysis.

In addition, the staff performed a oreliminary review of the licensee's operability evaluations. The opera)111ty evaluations were performed because the licensee determined that the containment fan coolers (CFCs) are potentially susceptible to SW boiling in the cooling coils (resulting in possible reduced flow and/or water hammer) during a design-basis LOCA with a concurrent LOOP and are susceptible to two-phase flow downstream of the CFC throttle valves during steady-state (after about 1 minute) post-accident conditions. The licensee's determination of this susceptibility was based on the Westinghouse letter NSAL 96-003. The licensee's operability evaluation was based on

' the work of two consultants, Fauske & Associates, Inc. (FAI) and Sargent

&-Lundy (S&L). As a result of that evaluation the licensee concluded that the SWS and CFCs would remain functional and continue to be operable based on interim operability criteria. The results of the licensee's analysis for water hamer showed that the bounding waterhamer loading would result in pipe stresses that exceed design allowables, but do not exceed the licensee's interim operability criteria. The interim operability criteria were determined by the licensee's comitments to IE Bulletin 79-14. " Seismic Analysis for As-Built Safety Related Piping Systems," and are based on American Society-l of Mechanicel Engineers (ASME) Section Ill, Appendix F limits. The staff performed a preliminary review of the licensee's interim i-operability evaluation including piping and piping support stress

analysis, As a result of that preliminary review the staff concluded that the licensee's approach in performing the oaerability evaluation was reasonable and_ appeared to be an acceptable ) asis for making;an interim operability determination until modifications are made during the next refueling outage.

?

results of the licensee's analysis for the potential delay in achieving full flow conditions showed that the maximum flow delay will occur in one of the lower CFCs (contrary to higher waterhammer loads in l

the-upper CFCs) because of the higher backpressure conditions that would exist comp 3 red-to the u)per coolers. The staff agrees with the licensee's assumption t1at the lower cooler flow assumed in the analysis will be the most limiting case for delayed achievement of full flow.

The revised containment heat removal analysis assumes a delay-of 62 seconds in CFC heat removal function. At 62 seconds, full heat removal capability is assumed with no credit for heat removal during the refill phase.

Based on at least three pumps running after 36 Teconds into the event the licensee has concluded that full flow will be achieved well within the 62 second time frame.

The staff has performed a preliminary review of the assumptions made for the associated two phase flow analysis and for the det.ermination of the'CFC refill velocities. As a result of its-review the staff has concluded that the licensee's assumptions for heat removal capability appear to be conservative and provide reasonable assurance that full CFC heat removal capability will-exist within the required 62 seconds.

In its September 30, 1996, submittal the licensee identified a number of possible modifications to eliminate the potential for water hamer and two-phase flow-in addition to possible modifications to make the water hammer and two-phase conditions part of the design basis for Point-Beach. The staff will perform a detailed review of the licensee's o)erability evaluation and proposed modifications during its review of tie licensee's overall response to GL 96-06. Additional actions may be required as a result of that review.

e.

TS 15.3.3.D.2 - The proposed change to TS 15.3.3.D.2.a would allow one SW pump to be-inoperable for u inoperable for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .p to 7 days, and two or three pumps to be The SWS is shared between the two units

' and the power arrangement is divided between the units in such a fashion that there are three electrical Train A powered pumps and three electrical Train B powered aumps. These six pumps are headered together and serve both units such t1at the loss of a pump or pumps affect both units.

This sharing also minimizes the effects from the 16ss of any one pump or any three pumps since cooling water flow will not be lost to any equipment of either unit. Therefore, none of the cooled equipment loses i

its cooling water supply even if all three of the inoperable pumps are

>owered from the same electrical train. As a result, the A0Ts do not lav9 to be associated with any particular electrical train.

With one inoperable pump the SWS still maintains a significant amount of redundancy with-the five remaining pumps. The basis for this TS change is that three SW pumps are necessary to mitigate the consequences of a l

design basis LOCA concurrent with a LOOP and shut down of the other unit. The staff considers 7 days as an acceptable A0T for one inoperable pump because of significant amount of redundancy left with the remaining five pumps and the low probability of a design-basis LOCA concurrent with a LOOP during the 7-day period. The 7 days also provides a sufficient amount of time to perform most repairs of these deep draft pumps that require pump / motor removal. Thus, the 7-day A0T.

provides a significant amount of flexibility without any significant reduction in plant safety.

During the proposed 72-hour A0T for two or three pumps inoperable the remaining 3 umps are adequate to perform all heat removal functions for all design.3 asis conditions including LOCA i

concurrent with a LOOP. However, the overall reliability is reduced because a single failure could result in loss of SWS function if (for the worst case) all three of the remaining operable pumps were supplied by the same diesel generator. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is consistent witn previous staff approved A0Ts for other ESF systems when an additional single failure can result in loss of all system function, The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on the redundant capabilities afforded by the remaining pumps, and the low probability of a design-basis accident during this time period.

It is also consistent with the Westinghouse Standard Technical Specifications. NUREG-1431.

The proposed change to TS 15.3.3.D.2.b allows the SW ring header continuous flow)ath to be out of service for up to 7 days and includes the provision tlat if fewer than four SW pumps are operaL'9. SWS flow shall be evaluated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after entry into this Lt0.

The current-specification specifies-that one of two loop headers may be out of service or-inoperable with an A01 of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The existing specification fails to recognize flexibility provided by the existing design and is not consistent with the actual design of the SWS. The SWS design actually is made up of three main supply headers identified as the north, south, and west headers (the SW pump comon discharge header can be visualized as the east header). The north and south headers each

- supply the west header during normal operation such that the overall effect is a continuous flowpath ring neader.

The proposed continuous flowpath LCO requirement would be applied any time continuity of the flowpath in the SW ring header is interrupted. This includes-isolation of any header or headers in the system.

This LC0 recognizes that one

aspect of redundancy in the SWS is the ability to isolate a break in the system and still maintain ability to provide required flow to supported equi) ment.

This capability is im) aired any time the continuous flowpath of tie ring header is blocked. Tie 7-day A0T is based on a piping failure occurring result'ing in system inoperability.

Piping failures' were considered as )assive failures and were not assumed to occur concurrent with a 03A.

(

The two main SW headers are designated " north" and " south." The south J

header 3redominantly serves Unit I components and the north header serves Jnit 2 components: additionally, each header also serves a number of shared components.

All shared safety features necessary for l

operation (EDG G-01 and G-02, motor-driven auxiliary feedwater pumps, CCW heat exchangers) can be supplied by either of these two main headers, each of which also su) plies the west header.

For Unit 1 two CFCs are supplied via the souti header and the other two CFCs are supplied via the west header.

For Unit 2. two CFCs are supplied by the north header and the other two CFCs are sup) lied by the west header.

The LCOs for the ecuipment rendered inoperaale by the header inoperability woulc be ap to another su) ply header. plied until the affected equipment is switched For unit-specific equi) ment' that cannot be supplied by tie other header (such as the CFCs) tie LC0 for that equi ament would be applied on that unit.

Accident mitigating capaaility. including the capability to meet the single active failure criterion. is maintained when the equipment is switched to another header.

The 7-day A0T is based on the fact that a passive failure (piping failure) must occur to cause a subsequent problem with system operability.

Depending on the segment of the SWS continuous loop that is isolated and the number and type of components affected, the total minimum SW flow required may be increased above that which could be supplied by three pumps. Therefore, the licensee has proposed a condition that SWS flow shall be evaluated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the ring header continuous flow)ath is out of service and less than four SW pumps are operable.

The 24 lours should provide adequate time to perform the required evaluation and to safely perform any necessary realignments.

If it is determined that any equipment will not receive sufficient flow, the applicable LCOs for the affected equipment shall be entered.

These LCOs can be exited if system realignment is completed to achieve the required flow rates for the affected equipment.

The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to complete the evaluation is the same as the existing specification for an inoperable header (loop) and is, therefore, considered acceptable to the staff.

The need for a 7-day A0T with a header out of service is based on potential cutting and welding repairs since a passive failure of a component or piping would require taking a header out cf service.

The 7-day A0T is acceptable because the staff does not postulate passive failures following DBAs and the SWS would still be capable of meeting the same active failure criteria with the ring header continuous flow path operable. Therefore, an A0T greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is warranted and

17 7 days is a conservative upper limit that provides flexibility for repairs without any significant decrease in plant safety.

The proposed changes to TS 15.3.3.D.2.c include more specific identification of the " valve" as ahy or all automatic isolation valves required during accident conditions thereby deleting the reference to passive components.

The changes also propose a 72-hour A0T for any out of service automatic isolation valves and would include a provision that at least four pumps must be operable for entry into this condition. The changes further propose that in addition to exiting the LCO by returning the valve to service, the LC0 may also be exited by isolating the affected line with a seismically qualified isolation valve or restoring the valve to operable status.

The change to only apply this LC0 to c

automatic isolation valves is acceptable since the existing specification could be interpreted to exclude these isolation valves because it applies to " valves or other passive components." Therefore, l

the existing specification appears not to apply to active valves. An LC0 for manual (passive) valves or other passive components is not necessary because the LC0 for the ring header would include any passive failure that affects the system flow.

The present specifications do not adequately address automatic isolation valves that are used to isolate nonessential equipment from the supply headers following an accident.

The failure to isolate this equipment could result in reduced flow to the accident heat loads and, therefore, a provision is provided to require four pumps to be o)erable for entry into this condition, The requirement for four operaale pumps assures adequate flow is available to mitigate DBAS.

Because the system cannot withstand all single active failures while 1 this-LCO. a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is appropriate and is consistent with the A0Ts for other ESF equipment and is consistent with NUREG-1431. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on the redundant capabilities of the remainder of the system and the low probability of a design-basis event during this time period.

The provision for exiting from this condition if the affected line is isolated by a seismically cualified isolation valve-is acceptable because, with the line isolatec. operation of the automatic isolation valve is no longer required.

A new LC0 is also proposed. TS 15.3.3.D 2.d. that allows the CFC outlet motor-operated valves to be o>en for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided at least five SW pumps are o)erable.

or a postulated accident in one unit, the outlet valves for tie affected unit receive full open signals (flow is limited through each cooler by individual manually operated SWS valves at the outlet of each cooler and upstream of these MOVs [ motor-operated valves]) while the unaffected unit's valves would remain closed. This LC0 is necessary to address testing requirements for the SW outlet valves. During testing each valve may be required to stroke full open.

Since this will increase flow through the full open valve. at least five SW pumps are required to be. operable.

If an accident is postulated to occur in the other unit (opposite to unit with the open valve) inadequate flow to the accident unit may result if a single active

-failure is also postulated.

Therefore, the "affected unit" would be the unit not being tested, i.e., the unit where the CFC outlet valves are

18 -

still closed.

If the outlet valves for both units are open, then both units are affected and the 72-hour A0T would-a) ply to both.

The provision for having at least five pumps opera)le to enter this LC0 is necessary to ensure adequate post-accident flow would be available. The proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months A0T is consistent with the A0Ts for the other SWS LCOs and is acceptable on the same basis.

Another 3roposed change to TS 15.3.3.D replaces the statement to allow one of tie conditions for TS 15.3.3.2.D.2 components to be ino>erable at any one time with a statement that effectively allows any of tie conditions to be in effect coincidentally. The change to allow coincidental entry into one or all of the SW LC0 conditions was accounted for by including the requirements that at least four or five SW pumps be operable under the specified LC0 conditions that would increase the SW flow requirements.

The additional flow required under the combination of these conditions is less than the capacity of a single SW pump except for Condition d, where the additional flow required could be more than the capacity of one pump but less than two pumps.

Therefore, the requirement for at least four pumps to be operable under Conditions b and c, and five pumps under Condition d.

provides assurance that the required flows are achieved.

In conclusion, since the provisions contained within the individual LCOs assure adequate flow capability is maintained for coincidental entry, the proposed change is acceptable.

If the requirements to restore any of the conditions within their specified A0T are not met, the licensee proposed to change the action statement to require only the affected reactor (s) to be shut down rather than both units. This change is acceptable since it recognizes that accident mitigation capability in only one unit may be affected and.

therefore,-provides flexibility to shut down only one unit if the other unit is not affected by the ino)erable equipment, The existing action statement always applied to bot 1 units and could result in unnecessary reactor shutdown transients if one of the units was not affected by the-out of service or inoperable equipment.

f.

TS 15.3.3 Basis - The staff has also reviewed the licensee's changes to-the Basis for TS 15.3.3 to include supporting information for each of

-the previously described changes. As a result of that review the staff concluded that the revised basis adequately supports the proposed TS changes and more accurately reflects the plant design and operation.

g.

TS 15.3.7.A.1.1. j. and k - The licensee proposed to change TS 15.3.7.A.1.1 which currently requires all normal and emergency power supplies to be operable only for the operation of both units (to be made critical). Under the proposed change, this LCO (i) would also apply to o>eration of only one unit. TSs 15.3.7.A.1.j. and k. currently specify tie normal and emergency power supplies that have to be operable for Unit 1, or Unit 2 to be made critical, respectively.

The change making LC01. apply to both units invalidates LCOs J. and k. such that they can no longer be applied. They were, therefore, necessarily deleted by the l

o

19 licensee's proposed changes. The proposed changes support the requirement for six operable SWS pumps, and are more conservative than the current specifications because all normal and emergency power sup) lies are now required to be operable for single unit operation ratier than just two unit operation. The proposed cha'nges are, therefore, acceptable.

h.

TS 15.3.7.B.1.g - This TS currently requires in part that if the normal

)ower sul) ply or standby emergency power supply to Unit 1 A06/B04 or Jnit 2 A)5/B03 is out of service, operation can continue for up to 7 days.

If this LC0 cannot be met-the affected unit will be ) laced in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown wit 11n 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The proposed change to TS 15.3.7.B.1.9 requires both units, rather than the affected unit, to be placed in hot shutdown within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months if the normal power sup)1y or emergency power supply to Unit 1 A06/B04 or Unit 2 A05/B03 or bot1 are out of service for more than 7 days. The proposed change is conservative in nature because both units must be shut down, and is siso necessary to support six operable SW pumps.

It is, therefore, acceptable.

i.

TS 15.3.12 and TS 15.4.11 - The licensee's new dose analysis assumed an increase in the control room emergency filtration (CREF) system charcoal adsorber filter efficiency from 90% to 95%. To ensure that the charcoal was actually capable of performing at the 95% level. the licensee proposed to change the acceptance criteria for the laboratory testing of charcoal in TS 15.3.12.1.b to a minimum allowable methyl iodide removal efficiency of 99%.

The licensee also proposed to change the TS basis to specify that representative CREF charcoal samples will be laboratory tested in accordance with American Society for Testing and Materials (ASTM) 03803-1989. " Standard Test Method for Nuclear-Grade Activated Carbon." with the test temperature-at 30 *C and the relative humidity at 95%.

The staff nas evaluated the above. proposed changes es contained in TS 15.3.12.

Because the CREF system charcoal has a depth of 2 inches, the charcoal is capable of an efficiency of 95%. The staff has concluded that an increase in the laboratory acce]tance criteria for allowable methyl iodide removal efficiency from tie present value of 90%

to 99% will provide adequate justification for assuming that the CREF system charcoal will perform at least at a level of 95% if called upon to mitigate the consecuences of an accident. The 3roposed acceptance criteria of 99% incluces a safety factor of 5'whici )rovides an acceptable degree of assurance that, at the end of tie operating cycle.

the charcoal will be capable of performing at a level at least as good as that assumed in the licensee's evaluation.

The staff finds these changes acceptable, j.

The proposed change to TS 15.5. " Design." Section 15.5.2.C..

" Containment Systems." Item 2., changes the heat removal capability for 1

r I

20 the containment vessel internal air recirculation system, four ventilation fans and air coolers, from a total heat removal capability of 55.000 Btu /sec to 41.700 Btu /sec under conditions following a LOCA.

This change is consistent with the throttling of the CFC service water outlet valves, and 'the analyses performed in support of this TS ' change

request, it is, therefore, acceptable.

3.2 Effect on Eauioment Qualification The licensee identified changes to the DBA conditions that are postulated based on information provided in support of the proposed changes in TSCR 192.

The changes in the DBA conditions required revisions to the DBA tem)erature and pressure profiles utilized in the environmental qualification (EQ) program. The licensee's initial evaluation and supporting information were discussed at a public meeting held March 3, 1997.

The majority of components evaluated were within the new DBA pressure and temperature profiles. The licensee's submittal of April 18, 1997, and June 3, 1997, provided additional clarifying information of the DBA conditions, an overall assessment of all components reviewed, and specific information on four components whose temperature profiles included in the original evaluation required additional documentation. The revised DBA analysis results in a peak tem 291 F which decreases to 278 F at approximately 7.5 seconos. perature of The revised temperature profile includes the saturated steam temperature based on the l

partial pressure of steam in containment. The saturation temperature is considered to be a more representative temperature arofile for equipment and structures inside the containment which would be su)jected to condensation heat transfer.

The licensee conclusions and bases for adequacy of the four components are described below:

a.

The E0 testing of the Transamerica Delaval Gems Level Transmitter envelops the new postulated PBNP DBA, except for a 5.8-hour )eriod.

This is judged to be acceptable because aging test data in t1e E0 test program envelops the points not met in the new postulated PBNP DBA. The Transamerica Delaval Gems Level Transmitter was tested to DBA conditions significantly harsher than those anticipated at PBNP except for a 5.8-hour aeriod (from 2.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months at 242 F to 7.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months at 225 F).

Prior to.0CA testing the transmitters were thermally aged at 248 F for 2,161.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

The demonstrated ability of the transmitter to survive extended periods at 248 F during thermal aging provides assurance that the transmitter will not fail during the 5.8-hour post-LOCA period when maximum temperatures of 242 *F (17 F above the EQ test data for the transmitter) may occur.

b.

The Rome cable is not required to be qualified because it performs no safety-related function and has no failure mode affecting safety for this DBA. Additionally, the licensee concluded that the failure of the Rome cable in any manner will not prevent any other safety-related equipment from functioning nor will it fail in any manner that could provide misleading information to the plant operators.

Therefore, the l

21 -

licensee concluded that the evaluation of the Rome cable to the design basis event LOCA conditions is not required based on the criteria specified in NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment." Revision 1.

c.

The E0 testing of the containment accident fan motor bearing lubricant (WCAP 7722. " Safety Related Research and Development for Westinghouse Pressurized Water Reactors Program Summaries." Spring 1971) envelops the new postulated PBNP DBA. except for a 5.25 hour2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months period at between 241 F and 220 F and an 8.5-hour period at between 188 F and 170 'F.

The licensee judged this to be acceptable because another E0 test re> ort (WCAP 7829. " Fan Cooler Motor Unit Test." A)ril 1972) envelops tie points not met in the new postulated PBNP D3A.

WCAP 7829 was a test of the containment accident fan motor, testing the Chevron BRB #2 lubricant at temperatures above 270 F for over 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .

The licensee's evaluation states that this testing proves that the containment accident fan motor bearing lubricant will not fail during the time when the E0 test data from WCAP 7722 does not envelop the new postulated DBA temperature profile. The current lubricant. Chevron SRI. is similar in composition to the Chevron BRB #2 and has a bearing life at 300 F and 10.000rpmontheorderof6.000hoursascomparedtotheoriginalBRB l

2 lubricant which has a bearing life at 300 F and 10.000 rpm on the order of 3000 hours0.0347 days 0.833 hours 0.00496 weeks 0.00114 months , d.

The containment accident fan has two lubricants. Chevron BRB #2 has been replaced by Chevron SRI.

The evaluation of the lubricant is the same as the evaluation performed for the containment accident fan motor bearings.

Westinghouse material M53701TT (E.I. DuPont deNemours & Co.

Inc. 's Krytox 240 AC Fluorinated Lubricating grease) is used in the 7

pillow block's labyrinth seal to help prevent chemical spray from L

reaching the roller bearing lubricant following initiation of containment spray.

Krytox 240 lubricant is rated for long periods at continuous temperatures up to 550 F and intermittent tem)eratures of 800 F.

The licensee judged both of these lubricants to 3e acceptable because another E0 test report envelops the points not met in the new postulated PBNP DBA for one lubricant and the other lubricant has a continuous use temperature rating enveloping the points not met in the new postulated PBN) DBA.

The staff evaluated the additional information regarding the four components which were determined to need further clarifications for their acceptability with respect to the new postulated conditions.

Based on engineering judgement, the staff agrees that there is reasonable expectation that the Transamerica Delaval Gems transmitters will survive the 5.8-hour post-LOCA i

l period at 242 F (17 F above the E0 qualification test) given that the transmitters were able to survive thermal aging testing at 248 F for over 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months . The staff agrees with the licensee that tnere is no requirement to qualify the Rome cable because it is not relied upon to perform a safety-related function.

The staff's evaluation of the lubrication data finds the information provided by the licensee is acceptable.

Therefore, the staff IEiLTA2.N 4

W

22

-\\

finds the licensee's review of the effects on EQ in response to the proposed TS change acceptable.

3.3 Evaluation of Post-Accident Samolina System The licensee evaluated the effect of the-revised containment pressure profile (Case 1) on the operation of the post-accident sampling system (PASS).

NUREG 0737.~ Item II.B.3 required the capability to promptly obtain containment atmosphere samples following a LOCA with the combined time allowed for sampling and anal to take a sample.ysis being 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months or less from the time a decision is made As discussed in the supplemental information dated June 3.

1997, the PASS system will be u) graded prior to restart of each unit-to allow a sample to be taken within an lour following accident initiation. This upgrade will reduce dose to personnel obtaining the sample to within GDC 19 limits. Since the modification is still in progress. Region III personnel l-will verify this commitment is completed prior to restart of each unit.

3.4 Evaluation of Radioloaical Consecuences of Desian-Bases Chanaes The licensee evaluated the radiological consequences of several DBAs including a large-break LOCA. steam generator tube rupture, control rod ejection.

reactor coolant pump locked rotor, and main steam line break.

The licensee determined that the radiological consequences are acceptable based on the above changes and the new steam generators. The licensee stated in its

~ application that calculated doses are below the acceptance criteria presented in 10 CFR Part 100.

The licensee's calculated doses are below the dose limits presented in GDC 19 for all of.the DBAs. analyzed provided that potassium iodide (KI) tablets are used by control room operators in the event of a

-large-break LOCA. control rod ejection accident, and locked rotor accident.

The licensee's calculations confirm that the LOCA doses are greater than the doses associated with the other evaluated DBAs.

In a-letter dated March 5.1997, the licensee committed to achieve compliance with GDC 19 dose limits without the use of KI tablets or respiratory

-protection devices.

The licensee has committed to submit new analyses and

--evaluations for complying with GDC 19 dose limits without the use of KI tablets or su) plied air breathing apparatus in sup) ort of'a license amendment request by Fe)ruary 27. 1998.

Implementation of tie proposed changes needed to comply with GDC 19 dose limits without reliance on KI or respiratory

. protection devices are expected to be completed within 2 years of the date

-that NRC approval for the license amendment is granted.

These commitments are included as additional conditions-to the licenses.

- The staff reviewed the licensee's analyses and compared the potential-radiological consequences to the current licensing basis and the dose limits presented in 10-CFR Part 100 and GDC 19 of Appendix A to 10 CFR Part 50.

' The.. staff also independently calculated the postulated radiological doses for individuals-located at the Exclusion Area Boundary-(EAB). Low-Population Zone (LPZ)..: and control room.

Provided below is-a-description of the assumptions that were used by the staff and the licensee for a large-break LOCA that was

reanalyzed based on the proposed TS changes and the proposed changes to system operation.

The licensee calculated the potential consequences of a postulated large-break LOCA to the control room operators and to individuals located at the EAB and LPZ. The sources of releases in the event of a LOCA include containinent leakage and ECCS recirculation loop leakhge.

Containment leakage sources released to the reactor building are assumed to be reduced by the effect of ESF equipment such as sprays and fan coolers.

Recirculation loo) leakage is assumed to be released with no credit for holdup or filtration

)y ESF equipment.

3.4.1 Containment Leakage Pathway In its application dated Se 30, 1996, and subsequent submittals, the licensee revised the large ptember break LOCA dose calculation to include an assumption that operation of the CS system would terminate before the end of the accident. The original large break LOCA analysis for plant licensing (1970) took credit for continuous operation of one CS and two CFCs throughout the accident.

The September 30, 1996, analysis assumed termination of CS after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months irito the accident and operation of two CFCs throughout the accident., The assumed time for terminating the CS system was revised twice in subsequent submittals, due to changes in the assumed RWST available volume resulting from the consideration of level instrument inaccuracies and net positive suction head calculations. The June 3.1997, submittal assumed termination of the CS system in 65 minutes, which is the value used by the NRC staff in the verification calculations.

The CS system operation was also reanalyzed with respect to the assumed spray removal coefficients for elemental and particulate-iodine.

The licensee used the methodology described in Revision 2 of Standard Review Plan (SRP) 6.5.2.

which is a methodology approved by the NRC. The calculated lambda for particulate iodine was 31.46 per hour, but a value of 20-per hour was used in the dose calculation per SRP recommendations. The revised lambda for particulate iodine of 6.02 per hour was found-to be acceptable also. The licensee calculated a limiting decontamination factor for elemental iodine of 340. This value is based on a partition coefficient of 10.000 from NUREG/CR-2900. (" Predicted Rates of Formation of Iodine Hydrolysis Species at

>H Levels. Concentrations, and Temperatures Anticipated in LWR Accidents").

rigure No. 33. assuming a pH of 7. temperature of 100 *C, and time of about R

one-half hour.

In accordance with SRP recommendations, the licensee used the maximum allowable decontamination factor of 200, which is a conservative assumption. The decontamination factor for particulate iodine was 19.8.

assuming that particulate iodine was removed only in a sprayed portion of the containment.

However, some mixing between sprayed and unsprayed containment volumes occurs, which was factored into the licensee's dose calculations.

24 3.4.2 ECCS Leakage Pathway The licensee's control room habitability analysis required by item 111.D.3.4 of NUREG 0737, dated September 4, 1984, assumed an ECCS leak rate of 25,000 cubic centimeters (cc) per hour, which is equivalent to 417 de per minute.

The licensee's June 3,1997, submittal assumed a decrease in ECCS leak rate outside containment to 400 cc 3er minute for equipment leakage from the ECCS

-systems outside containment.

or the purpose of c61culating radiological doses for a design-basis LOCA, the staff normally considers the equipment leakage rate of all components in the ECCS system.

The licensee calculated doses from the ECCS equipment leakage pathway at the '

EAB, LPZ, and control room.

For the EAB and LPZ dose calculations, the licensee assumed double the ECCS leak rate. The licensee did not double the i

ECCS leak rate for control room dose calculations.

The LOCA dose calculations assumed a volume of water in the containment sump of 197,000 gallons, The auxiliary building filtration system was not credited for iodine removal from ECCS leakage in the licensee's analysis.

The licensee's new dose analysis assumed an increase in the CREF system charcoal adsorber filter efficiency from 90% to 95%.

3.4.3-Additional Assurance In the letter of June 13,1997 (NPL-97-0351), WEPC0 submitted specific actions to provide additional assurance that the assumptions made in the dose calculations would not be exceeded.

These actions include (1) performance of monthly leakage inspections of accessible portions of the ECCS outside containment that could contain recirculated fluid from the containment during a LOCA, (2) inspecting accessible, pressurized ECCS piping outside containment during quarterly inservice testing, (3) performance of the leakage reduction and preventive maintenance program tests for the ECCS during any cold shutdown outage of sufficient duration (about-5 days or longer) in which 6 months or more has elapsed since the previous testing (4) performance of corrective action based on the results of these inspections and-tests to ensure ECCS leakage remains as low as reasonably achievable. (5) performance of periodic inspections of the control room ventilation system to. verify adequacy of material condition,- and (6)-increased testing ~ of the control room ventilation system filters to approximately 6-month intervals. The staff relied upon-

- these s)ecific actions to provide a reasonaole assurance that assumptions used 1

-as the ) asis of the radiological consequences are maintained.

3.4.4 Sumary of Radiological Consequences Assessment Based on the assumptions provided by the licensee, the staff's evaluation indicates that termination of the-CS system 65 minutes after initiation results in radiological doses at the EAB and LPZ that are within the acceptance criteria presented in SRP 15.6.5, ' Appendices A and B of NUREG-0800.

Radiological doses to the control room operators are within the acceptance criteria of SRP 6.4 of NUREG-0800,- based on the licensee's commitment that

25 control room operators will utilize K1 tablets in the event of a large break LOCA. Additional details on the assumptions for this evaluation are ) resented in Table 1.

The-staff assessed the upon the assumptions in this tabl,e, potential consequences of a LOCA 3ased The thyroid doses are presented in Table 2, Based on the staff's review, licensee comitments, and the licensee's decision to provide additional assurance above those required by the proposed TS, the staff has concluded that there is reasonable assurance that the dose limits presented in 10 CFR Part 100 and GDC 19 would not be exceeded. Therefore, the staff considers the licensee's analyses acceptable.

3.5 Evaluation Sumary The new analyses indicate that (a) containment post-LOCA long-term cooling is adequate to prevent containment overaressure and (b) the post-accident containment pressure response upon w11ch containment leakage and associated l

radiological dose calculations are based remains valid.

3,6 Tvocarachical Correction A typographical error on TS page 15.3.12-1 has been corrected by adding the number "2." to the items listed under Soecification. This number was inadvertently omitted when the staff issued Amendments 110 and 113.

4.0 EXIGENT CIRCUMSTANCES

The Comission's regulations,10 CFR 50.91, contain provisions for issuance of amendments where the Commission finds that exigent circumstances exist, in that a licensee and.the Comission must act quickly and that time does not permit the Comission to publish a Federal Register notice allowing 30 days for prior public coment. The exigency exists in that the aroposed amendments are needed prior to the restart of Unit 2 which relies on slared systems that require the issuance of the amendment for Unit 1 concurrently and time does not permit the Comission to publish a notice allowing 30 days for prior public coment.

The licensee states that the circumstances of exigency were not avoidable, based on the need to refine and revise the submittals due to emergent issues.

The licensee revised its original application on November 26 and December 12, 1996. February 13, March 5 April 2 April 16. May 9. June 3.

June 13 (two), and June 25, 1997. At the time of the licensee's June 3,1997, submittal, Unit 2 restart was scheduled for July 1,1997. The staff has determined that the licensee used its best efforts to make a timely application.

Accordingly, the Commission has determined that exigent circumstances exist pursua1t to 10 CFR 50.91(a)(6), the submittal of information was timely and the exigency could not have been avoided, and that the licensee did not create the exigency,

26 -

5,0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATIONS DETERMINATION The Commission's regulations in 10 CFR 50.92(c) state that the Commission may make a final determination that a license amendment involves no significant haza'ds consideration if operation of the facility in accordance with the r

pro)osed amendment would not (1) involve a significant increase in the pro) ability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) result in a significant reduction in the margin of safety. The NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendments and that the amendments should be issued as allowed by the criteria contained in 10 CFR 50.91. The NRC staff's final determination is presented below.

(1)

The proposed changes would not involve a signific6nt increase in the probability or consequences of an accident previously evaluated.

The probabilities of accidents previously evaluated are based on the probability of initiating events for these accidents.

Initiating events for accidents previously evaluated for Point Beach include control rod l

withdrawal and drop, CVCS [ chemical volume and control system]

l malfunction (boron dilution), startup of an inactive reactor coolant l

loop, reduction in feedwater enthalpy, excessive load increase, losses of reactor coolant flow, loss of external electrical load, loss of normal feedwater, loss of all AC power to the auxiliaries, turbine overspeed, fuel handling accidents, accidental releases of waste liquid or gas, steam generator tube rupture, steam pipe rupture, control rod ejection, and primary coolant system ruptures.

This license amendment request proposes to change the limiting conditions for operation, action statements, allowable outage times, and design specifications for the PBNP TS associated with the containment accident fan coolers, SW equipment, CCW system, control room ventilation system, and normal and emergency power supplies.

These proposed changes do not cause an increase in the probabilities of any accidents previously evaluated because these changes will not cause an increase in the probability of any initiating events for accidents previously evaluated.

In particular, these changes affect accident mitigation systems and equipment which do not cause accidents.

The consequences of the accidents previously evaluated in the PBNP FSAR

[ final safety analysis report) are determined by the results of analyses that are based on initial conditions of the plant, the type of accident, transient response of the plant, and the operation and failure of equipment and systems. The changes proposed in this license amendment request provide appropriate limiting conditions for operation, action statements, and allowable outage times for SW, CCW, containment cooling, control room ventilation filtration, and normal and emergency power supplies.

27 -

The proposed changes affect components that are required to ensure the proper operation of ESF equipment. The proposed changes do not increase the probability of failure of this equipment or its ability to operate as required for the accidents previously evaluated in the PBNP FSAR.

The proposed changes that increase the allowed outage times for ESF equipment continue to provide appropriate limitations for these conditions because sufficient equipment is still required to be operable for accident mitigation and the proposed allowed outage times are consistent with currently accepted tina periods for these situations.

Therefore, these proposed license amendments do not affect the consequences of any accident previously evaluated in the PBNP FSAR, because the factors that are used to determine the consequences of accidents are not being changed.

(2)

The proposed changes would not create the possibility of a new or different kind of accident from any accident previously evaluated, New or different kinds of accidents can only be created by new or different accident initiators or sequences.

New and different types of accidents (different from those that were originally analyzed for Point Beach) have been evaluated and incorporated into the licensing basis for PBNP.

Exampics of different accidents that have been incorporated into the Point Beach licensing basis include anticipated transients without scram and station blackout.

The changes proposed by this license amendment request do not create any new or different accident initiators or sequences because these changes to limiting conditions for operation, action statements, allowable outage times, and design specifications for SW, CCW, containment cooling, control room ventilation filtration, and normal and emergency po - ]plieswillnotcausefailuresofequipmentoraccident L

different than the accidents previously evaluated. Therefore,

.o t' o - r med TS changes do not create the possibility of an accident it type than any previously evaluated in the Point Beach 01 H

FSA.

(3)

The propo m changes would not result in a significant reduction in the margin of safety.

The proposed changes provide the appropriate limiting conditions for operation, action statements, allowable outage times, and design specifications for SW containment cooling, CCW, control room ventilation system, and normal and emergency power supplies. This ensures that the safety systems that protect the reactor and containment will operate as required. The design and operation of the reactor and containment are not affected by these proposed changes. The proposed changes resulted in revised design bases for both units. The revised design bases were appropriately evaluated to ensure that there was not a significant reduction in the margin of safety. The safety systems and limiting conditions for operation for these safety systems that provide

o A support functions will continue to meet the requirements for accident mitigation for PBNP.

6.0--STATE CONSULTATION In accordance with the Comission's regulations, the Wisconsin State official

- was notified of the proposed issuance of the amendments.

The State official had no coments.

7.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the federal Regfster on July 9, 1997 (62 FR 36852).

Accordingly based upon the environmental assessment, the' Commission has determined that the proposed action will not have a significant effect on the

-quality of the human environment.

8.0_

CONCLUS10ti The staff 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 pro)osed manner, (2) such activities will be conducted in compliance wit 1 the Comission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

W. LeFave W.-Long S. Saba K. Manoly K. Parczewski R. Emch Date: July 9.-1997 -

Table 1 Licensee's Assumptions for LOCA Analysis Core Thermal Power (HWt) 1549 Activity Released to the Reactor Building Airborne (fraction of core) lodine 0.5 Noble Gases 1.0 Iodine Plateout Factor 0.5 Iodine Species (fraction)

Elemental 0.91 Particulate 0.05 Organic 0.04 Activity Released to Sump (fraction)

Iodine 0.5 l

Noble Gases 0.0 i

Containment Building Free Volume (ft')

1.065E6 Leakage Rate (%/ day) 0-24 hours 0,4

> 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0.2 Sump Liquid Volume (gal) 1.97E5 Containment Cooling Unit Number of Fans 2

Total Flow Rate (cfm) 7.70E4 Recirculation Efficiency (%)

0 Reactor Building Spray System Actuation Time (sec) 90 Duration Time (min) 65 Spray Removal Constants (/hr)

Elemental 20 Particulate 6.02 Fraction of Reactor Building Sprayed 0.45 Fraction of Reactor Building Unsprayed 0.55

o Table-1

-Licensee's-Assumptions for.LOCA Analysis (continued)

Recirculation Loop Leakage Rate for EAB and LPZ (cc/ min).

800 Leakage Rate for Control Room (cc/ min) 400 Gross Failure of a Passive Component 0

Minimum Time to Recirculation (min) 20 3

Atmospheric Dispersion Factors (sec/m )

[

EAB 5.0E-4 LPZ 0 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.0E-5 8 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.6E-5 1-4 days 4.2E-6 4-30 days 8.6E-7 Containment Leakage to Control Room 0 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

3. 0E-3 8-24 hours 1.9E-3 1-4 days 1.2E-3 4-30 days 4.8E-4 ECCS Leakage to Control Room 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1.7E-3 8 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

-1.2E -

.1-4 days 6.7E-4 4-30 days 2.3E-4 Breathing Ratec (m /sec)

Offsite 0-8 hours =

3.47E-4 1

8-24 hours-1.75E-4 =-

1-30 days 2.32E-4 Control Room 3.47E-4 1

i

-w.

---4 x.

o Table 1 Licensee's Assumptions for LOCA Analysis (continued)

Control Room Free Volume (ft')

65.243 Mode 4 Makeup Air Filtration Rate (cfm) 4,950 Mode 4 Unfiltered Inleakage (cfm) 10 Mode 4 Makeup Filtration Efficiency (%)

Elemental 95

" articulate 99 Organic 95 Occupancy Factors 0-1 day 1.0 1-4 days 0.6 4-30 days 0.4 i

.o

v Table -2 LICENSEE'S CALCULATED THYROID DOSES FOR POINT BEACH UNITS 1 AND 2 LOSS OF COOLANT ACCIDENT DOSE (rem)-

4 LOCATION-Containment Leakage:

ECCS-Total Leakage 2

EAB 133.3 57.2 190.42 1

LPZ 24.37 37.0 61.37 Control Room 186.0 106.7 292.7 3 2

1.

NUREG-0800 Acceptance Criterion = 300 rem thyroid.

2.

NUREG-0800 Acceptance Criterion - 30 rem thyroid.

3.

The licensee assumes that administration of 100 mg potassium iodide tablets will reduce calculated thyroid doses by a factor of 10.

4 4

+

4

.

ML20217N374

Contents

Text

2.0 BACKGROUND

4.0 EXIGENT CIRCUMSTANCES

7.0 ENVIRONMENTAL CONSIDERATION

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ML20217N374

Text

2.0 BACKGROUND

4.0 EXIGENT CIRCUMSTANCES

7.0 ENVIRONMENTAL CONSIDERATION

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