ML20198T361

Responds to 971010 RAI Re License Amend Request 218.Increase in Mass Release & Consequences for Changes in Letdown Line Break Transient Do Not Warrant Consideration of Change in Es Actuation Setpoint

ML20198T361
Person / Time
Site:	Crystal River
Issue date:	11/07/1997
From:	Holden J FLORIDA POWER CORP.
To:	NRC
References
3F1197-26, TAC-M99571, NUDOCS 9711140265
Download: ML20198T361 (12)

Text

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C 0 R P O A AT 40 N November 7,1997 3F1197-26

Subject:

Response to Request for Additional Information - License Amendment Request 218 Revision of the Makeup System Letdown Line Failure Accident Analysis (TAC No. M99571)

Refert.nces: 1. FPC to NRC letter 3F0997-14 dated September 9,1997, License Amendment Request (LAR) 218

2. NRC to FPC letter dated October 10,1997, Request for Additional Information for License Amendment Request No. 218

3. FPC to NRC letter 3F0997-30 dated September 25,1997, Supplement to Technical Specification Change Request Notice 210 Dest Sir:

The following information is provided in response to the Reference 2 Request for Additional Information (RAI). The format of this letter reiterates the NRC staff questions provided in Reference 2 followed by Florida Power Corporation's (FPC) response. Additional questions raised during a telephone conference conducted with staff reviewers on October 14,1997, are included as item 9.

1. Based on our review of various Crystal River Unit 3 documents our understanding of the control complex habitability envelope (CCHE) configuration and operation is tabulated below. Please confirm these data:

Normal operations prior to event:

AliF-19A(B) 42,335 cfm -

AIIF-17A(B)' 48,000 cfm / Q l' ' /

AIIF-21 A(B) 4,805 cfm AIIF-20A(B) 10,655 cfm AIIF-30 5,390 cfm D-12 adjusted for 5,260 cfm D-1/D-ID adjusted for 5,700 cfm D-2 shut lf!!$ll$fl$f,lhlll]$fll~

D3 open (passing 42,335 cfm)

Crptal River Energy Complex: 15760 West Power Line Street + Crystal Rwer, FL 34428 6708 e (352) 795 6486 A florida Wss Company 9711140265 971107 PDR ADOCK 05000302 l P PDR.,,

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x ' ' U.S. Nuclear Regulatory Commission 3Fil97 26 Page 2 After recirculation mode actuation

~AliF-19A(B) 42,335 cfm AIIF-17A(B) 48,000 cfm (stopped manually after 30 minute delay)

AlIF-18A(B) 43,500 cfm (started manually after 30 minute delay)

AIIF-21 A(B) stopped ,

AlIF-20A(B) stopped-AIIF-30 stopped D-12 shut D-1/D ID shut

-D 2 shut <

D-3 open (passing 42,335 cfm)

FPC Responsei Several modifications are being made to the control complex habitability envelope (CCllE) and control complex ventilation system (CREVS) during the current Crystal River Unit 3 (CR-3) outage and affect the configuration as outlined above. These modifict.tions are associated with FPC Restart issue R-12 to improve the integrity of the CCilE and thereby the level of protection it provides. Generally, the modifications are aimed at decreasing inleakage through individual boundary elements. The following modifications are being implemented:

• Damper AllD-99, which brings outside air to the Ventilation Equipment Room, is being removed and a permanent blank installed. New supply and return registers are being installed in the ductwork which will now serve the

. ventilation for this area. Thb will eliminate AIID-99 as a potential source of inleakage.

• Existing damper AllD-12, weated in the supply duct to the Chemical Laboratory, is being removed and replaced with two new bubble tight dampers, AllD-12 and AIID-12D, to be installed in series.

e Existing damper AllD-2, located in the exhaust duct to the outside, is being locked open and abandoned in place. Two new bubble tight dampers, AHD-2C and AllD-2E, are being instalied in series. AHD-2C will be normally closed.

• The normal position of recirculation air damper AHD-3 will be determined '

upon completion of post-modification functional testing. Damper AllD-3 will pass a minimum of 37,800 cfm in the recirculation mode after consideration of fouled filter conditions.

• Dampers AHD-1 and AHD-ID, located in the air intake duct, are being disabled and abandoned in place. Two new bubble tight dampers, AHD-1C and AIID lE, are being installed in series on the inlet duct. Dampers AHD-IC, AHD-2C, and AHD-3 will have positioners to preserve a manual override feature. This will allow operators to position these dampers to modulate the outside airflow, as required.

• Mechanical Equipment Room Air Handling Fans, AHF-21 A/B and associated dampers AHD-24, AHD-25,- AHD-26, and AHD-27 are being spared in place

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. U.S. Nuclear Regulatory Commission 3F1197-26 Page 3 and the associated CCilE penetration fitted and sealed with a seismic cap. This portion of the system previously exhausted air from the Mechanical Equipment i Room, Elevator Equipment Room, lavatory, kitchen, and toilet. This eliminates another potential source of inleakage into the CCllE.

• New supply and return registers are being installed in the ductwork in the Ventilation Equipment Room. This will provide full-time ventilation to this portion of the CCIIE during both normal and recirculation modes, e A skid-mounted air handling unit consisting of a fan and a charcoal filtration unit will be installed to ventilate the Elevator Equipment Room, lavatory, kitchen, and toilet. This system is non-safety and non-seismic, does not communicate witk the outside atmosphere, and will vent approximately 1,000 cfm by way of a non-safety related duct.

• Drain piping penetrating the CCIIE is being fitted with loop seals e orevent inleakage through the lines.

. Vestibules have been installed over all CCllE boundary doors, and have been sealed to provide maximum leaktightness. These vestibule: provide a means to test individual CCIIE boundary door leaktightness as well as reducing inleakage associated with CCIIE access / egress.

The post-modification functional testing associated with FPC Restart issue R-12, performed upon completion of the above modifications, will assure overall balancing and proper How rates for the Control Complex Ventilation System. The configuration and mod I for the control room dose calculations will be described in a separate submittal associated with the control complex habitability envelope (CCIIE), currently scheduled to be sent to the NRC on November 10, 1997.

2. For the letdown line event, at what time would control room automatic isolation and shift to recirculation occur? If automatic isolation does not occur (same signal that would have isolated le'down?), what is a conservative estimate of time for operator to actuate this shift to recirculation? (i.e., what is duration of unfiltered 5700 cfm intake after start of letdown line release but prior to isolation?)

FPC Response:

Initiation of control room isolation occurs at the same time as isolation of the letdown line. The Manual RBIC signal discussed in Reference I will isolate the control complex and is the same signal that isolates letdown. The credited time for isolation of letdown using manual actuation of reactor building isolation and cooling (RBIC) is ten minutes, consistent with NRC letter to FPC dated September 26,1978, which presented the staff position regarding allowable operator actions for which credit may be taken following a Condition 111 event (small LOCA).

The Maximum liypothetical Accident (MIIA), loss of coolant accident / loss of offsite power, is assumed to initiate control room isolation by the engineered safeguards (ES)

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-- signal associ ti with the 4 psi containment building pressure. J A review was Jconducted ot ents which might require Control Complex Radiation Monitor RM-A5_-  !

to initiate iso ion of the control complex habitability envelope. Based upon this' j

review, the St. 'n Generator Tube Ruptune (SGTR) accident was selected as the--

limiting event vy which to determine the impact of radiation monitor response on:

Control Room dose (see answer to Question 6 for additional information); This .  ;

analysis showed that the SGTR accident r wid not result in operator dose in excess of:  !

regulatory limits.-

' Approved vendor calculations, currently under review by FPC, show that the SGTR -

will result in Control Operator thyroid doses well below the limits imposed by GDC 19_ ,

even without isolation of the control complex ventilation system by RM A5. Radiation -

Evalues sufficient to alarm the radiation monitor are assumed to occur approximately 13  :

- seconds into the event based on conservative assumptions regarding monitor response.- ,

Assuming any reasonable isolation time either by the radiation monitor or manually by the operator, the MilA remains the limiting event with regard to operator dose.

3;' If automatic isolation and shift to recirculation is based on recirculation line radiation monitor alarm, provide basis of alarm response, e.g., equivalent isotopic concentration in duct (particularly that for iodine). Include any expected delays, e.g., damper movement; diesel sequencing of dampers, monitors.

FPC Rc;ponse:

. RM-A5 is located in the exhaust duct of the control room. It monitors the control room vent return duct for gaseous and iodine activity. It consists of iodine and gas-measuring channels preceded by a particulate filter. Channel checks, channel functional tests, and channei calibrations are performed on RM-A5 in accordance with Technical Specification 3.3.16 surveillance requirements.

The analysis for a bounding event, SGTR, was done for Xe-133 contribution only.

This is considered conservative for two reasons. First, by excluding all other isotopes,

. concentration of activity in the control room is effectively reduced. This in turn increases the time to trip the alarm and hence, is more conservative. The second

- reason that the Xe-133 concentration yields a conservative result is all the other noble gases released produce high energy beta radiation.-: The high energy beta constitute a

. large portion of the radiation that, in reality, will be counted by the detector. _ By excluding them in the' analysis, the' actuation time is conservative. Another assumption is that the alarm will trip on the noble gas long before it trips on the iodine. Since the -

chances of large quantities of elemental iodine being released in a gaseous phase are very 9 mall, it'is logical to conclude that the monitor will alarm on noble gags first.

- ' RM-A5 is fed from an inverter backed, vital bus. The' design is such that there would be no significant delays in the signal from RM-AS. The dampers are designed to stroke full closed within 5 seconds of the RM-A5 actuation signal. The dampers are e

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(aligned to the recirculation position by de-energizing a solenoid valve and are not

? dependent on diesel sequencing.

- 4. - . Connrm that'the mass release estimate of 114,000 ' Ibm from the letdown line is based on the total time to leakage isolation at 1170 seconds.

FPC Response:

The 114,000 lbm release begins when the break occurs at time zero, and continues until 1170 seconds (10 minutes after the hot legs reach saturation conditions) it which time the operator is assumed to have isolated the break by closing the containment isolation valves on the letdown line. This is discussed in the first paragraph under Section 14.2;2.6.4 ("Results of Analysis") of the revised Final Safety Analysis Report

- (FSAR) text provided with Reference 1 and is supported by FPC Calculation MM-0043, "CR-3 RELAPS Letdown Line Break."

- 5. _ Provide justification for not assuming that the system depressurization associated with the letdown line failure results in an incident-induced iodine spike.

FPC Response:

' The reactor coolant system (RCS) is assumed to depressurize as a result of the break.

Due to the automatic increase in makeup flow on low pressurizer level and nashing of the reactor coolant once pressure decreases to saturation for the hot legs, the RCS pressure does not decrease to the high pressure injection (HPI) actuation setpoint.

The dose analysis does not consider iodine spiking. CR-3 was originally licensed to the' Atomic Energy Commission " A Guide for the Organization and Contents of Safety Analysis Reports" dated June 30,1966, which did not include the letdown line, rupture accident. FPC chose to consider including the letdown line failure accident in February,1979 since Regulatory Guide 1.70 Revision 3 had been issued in November 1978 and suggested an analysis of a failure of small lines carrying primary coolant outside containment. The dose consequences performed for the letdown line failur:

accident were consistent with those previously approved methodologies reported in the (FSAR. The dose consequences were not performed in accordance 'with the Regulatory l Guide but were performed in accordance with the design basis for CR-3 which does not consider accident-irxiuced iodine spiking. Further, in a letter dated March 31,1983, FPC provided the NRC with a request for a technical specification change in which a statement was contained that each FSAR accident analysis had been examined with respect to changes in Cycle 5 parameters. The SGTR accident dose consequence values calculated for the reload report were performed "without consideration of iodine spiking." -In NRC letter dated July 12,1983 which provided the safety evaluation.

report for Amendment No. 64 to FPC's license, the NRC stated a comparison of the radiological doses calculated for Cycle 5 to those previously reported for Cycle 3 -

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'shows that all dose values are either bounded by the Cycle 3 values or are a small-

'  ; fraction of the 10 CFR 100 limits.

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Submittal does not address the impact of the proposed change on the control room --

- 6. [

- habitability. Please discuss the impact of the letdown line failure on the control -

complex habitability to satisfy your final safety analysis report (FSAR) Criterion 11'  ;

_(General Design Criterion 19). Please address impact of differences in source mix,

- source magnitude (including iodine spike) and accident progression on automatic and  ;

-- manual control room isolation actions; FPC Response:

l The Letdown Line Failure Accident was not analyzed with respect to Contr'ol Room

_ Operator dose. _However, as noted above, a review was conducted of events which might rely upon Radiation Monitor RM-A5 for isolation to determine the impact that radiation monitor response could have on operator dose and to verify that the MHA remains the limiting event in this respect. No consideration was given to iodine spiking. <

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The limiting control room (CR) dose is based on the MHA and assumes a Regulatory-

. Guide 1.4 source term. The projected dose is 26.5 REM to the thyroid for the 30 day

- period following the accident. Both whole body and skin dose are much lower than the 10 CFR 100 limits. ,

FSAR Table 14-30 lists the amounts of various nuclides released to the atmosphere

' from a SGTR and resulting dose at the exclusion area boundary and low population zone (LPZ). Tables 14-42 and 14-43 contain the same data for the letdown line rupture.-

The SGTR offsite doses are either larger than or comparable to the letdown line rupture. The 30 day LPZ dose values for SGTR and letdown line rupture are 0.017 and 0.027 REM, respectively. By comparison, the MHA 30 day LPZ dose is 23.64 REM for thyroid as shown in FSAR Table 14-54. Therefore, the control room dose

- for SGTR or letdown line, as analyzed, do not exceed the MHA dose.

Considering the NRC question on longer letdown line isolation times, its effect on CR-13 dose is not linear. The majority of dose would result from radioactive material taken into the contrbl complex prior to isolation of the habitability envelope. . Once isolation occurs, a continued leak would have a much smaller additional contribution to dose.

c7i - Short-term accident. dispersion values are tabulated in FSAR Tables 2-18,14-23, and 14-52.-~ The values in Table 14-23 differ in magnitude (lower) from those in the other tables. ; Which values 'are currently applicable to design basis calculations at Crystal -

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- Page 7 FPC Response:

The offsite dose analyses for the FSAR Chapter 14 accidents are currently based on two sets of atmospheric dispersion (X/Q) values, as shown in Table 1. The offsite dose analyses for the Fuel llandling Accident, Control Rod Ejection, LOCA, and 12tdown Line Break are based on the X/Q values shown in FSAR Table 2-18. The offsite dose analyses for the Load Rejection, Station Blackout, Steam Line Break, SGTR, and Waste Gas Tank Rupture are based on the X/Q values show in FSAR Table 14-23.

The X/Q values listed in FSAR Teble 14-52 (for the design basis LOCA dose analysis) are consistent with FSAR Table 2-18.

TABIE1 X/O Valnec Used in FSAR Chanter 14 Offsite Dose Analyses Didance IJsle FSAR Table 2-18 FSAR 'I able 14-23 Fuel Handling Accident Load Rejection Control Rod Ejection Station Blackout LOCA Steam Line Break Letdown Line Break SGTR Waste Gas Tank Rupture EAB 0-I hr 2.1x10* 1.78x10*

O - 2 hrs 1.6x10* 1.55x10*

O - 8 hrs 1.4x10" 5.88x10*

LPZ 8 - 24 hrs 1.5x10* 3.81x10*

1 - 4 days 7.7x10 l.69x10*

4 - 30 days 4.5 x10 l.03x10*

8.- We understand that volume of the CCHE is 355,311 cubic feet, of which,85,573 cubic feet is the actual control room. Please:

(a), provide the volume of the area supplied by damper D-12, and AHF-30 and exhausted by AHF-20A/B, and the volume of the equipment room exhausted by AHF-21A/B.

(b) confirm whether they are free volumes (c) confirm whether the CCHE and control room volumes specified above include volumes discussed in (a) above.

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FPC Response:

The CCHE is the portion of the Control Complex which is automatically isolated by the CREVS boundary dampers upon receipt of an ES, toxic gas or radiation monitor ,

alarm, and credited with providing protection to the operator during these events. The

. . volume of the CCHE was recently recalculated as being 364,922 ft', a slight increase from the previous value of 355,311 ft'. This value excludes all portions of the Contro!

Complex outside the.CREVS botmdary dampers, including that portion on the 95' elevation supplied by- AHD 12 and AHF-30. The mechanical equipment room on the 164' elevation is within the CCHE boundary and is considered to communicate with the CCIIE. Modifications currently being implemented will provide full-time ventilation to this area during both. normal and emergency modes of operation. The volume of the Control Room elevation was recently recalculated as being 87,729 ft', slightly larger than the previously assumed value of 85,573 ft'.

9. Additional questions raised during October 14,1997 conference call by Human Factors Branch reviewer:

L A.

What step in EOP-3 causes the operator to isolate letdown? The step which closes the letdown line valves primarily deals with assurinFfull HPI and is described in EOP-13, Rule 1. Is this the right step?

FPC Response:

-- Letdown is isolated by implementation of Rule 1 of EOP-13 when loss of subcooling margin (SCM) is indicated. The " rules" within EOP-13 assume higher priority than-other EOPs with the exception of immediate actions rcquired 'oy EOP-2, " Vital System Status Verification." EOP-13 Rule 1, " Loss of SCM," requires actuation of manual RBIC which closes the letdown containment isolation valves. EOP-3, Step 3.4 also verifies letdown is isolated. Thus, isolation of letdown is assured during a loss of subcooling margin. Isolating letdown is very simple. There are two push buttons on the ES section of the main control board (MCB) to actuate manual RBIC. Each push button closes one train of containment isolation valves. Thus, depressing just one of the two push buttons will isolate the letdown flow path. Rule 1 requires both push buttons to be

. depressed. These steps have not changed from the proposed revisions discussed in Reference 3.

B. Where did the 10 minutes to isolate come from? Is this supported by the results from training the operating crews? If so, how many crews have been trained?

What is ycur confidence level that this assumption can be achieved?

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FPC Response:

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The break is ass 0thed to be isolated within 10 minutes after the hot legs reach

• saturation conditions by closing the containment isolation valves on the letdown line.

• This is accomplished manually by the operator for the letdown line rupture event by 7

. actuating RBIC. The EOPs call for RBIC actuation upon a loss of adequate subcooling

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margin. The 10 minutes assumed is a credited action consistent with the time for a -

simple operator action that can be accomplished from the control room.-

. 'All six operating crews plus backups have been trained as of October 24,1997. -The-operating crews have consistently performed the required actions on the plant-specific l simulator within the 10 minutes. The_ table summary of simulator validations provided as Attachment E in Reference 3 included a scenario for the letdown line rupture downstream ofletdown isolation valve, MUV-49. The table provided timing from the EOP validations but was primarily focused on tripping the reactor coolant pemps  ;

(RCPs) and did not specincally address the HPI and RBIC actuations. A simulator run was performed on October 17,1997 for a decay heat drop line break scenario. The scenario included LSM and the steps were timed. The RBIC actuation was performed .

on both "A" and "B" sides within~ 72 seconds of loss of subcooling margin. This is F representative of crew response. .

C, The entry conditions for EOP4 are Modes 3 and 4. What if the condition (loss of subcooling margin) arises in Mode 1 or 2? What is the applicable procedure i-then?

FPC Response:

EOP-3 is a Mode 3 and 4 procedure because it is written with the assumption that the reactor is suberitical prior to taking any other operator actions. If the transient occurred in Mode 1 or 2, the reactor would be tripped by the reactor protection system (RPS) on low pressure, or by operator action due to low pressurizer level (100"). EOP-2 is the.-

highest priority procedure. Per Al-505 " Conduct Of Operations During Abnormal and -

Emergency Events," operators will perform the first five immediate actions of EOP-2 and

.then scan for symptoms. Since loss of adequate subcooling margin is the next highest

'1 symptom, EOP-3 would be entered.; Also, the operators are required to memorize the applicability of the EOP Rules contained in EOP-13. Operators have laminated copies i of the rules-in pockets situated ir. front of the main control board. Operators will trip X . the four RCPs and actuate manual HPI and RBIC when adequate SCM is lost.

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D .- Is the dose analysis based on a release for 10 minutes? How does FPC know this release duration is acceptable? What is the impact if the release is longer

than 10 minutes or the' operator fails to isolate letdown entirely? What is FPC's l basis _for the words in the "no signincant hazards consideration" on page 9 of a__ a. .

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.Page'10H 1 Attachment A of Reference 17 Discuss what will happen if operator action is

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FPC Response:~

. The "No Significant Hazards Consideration" on page 9 of Attachment A'of Reference 1 states that " adequate time'would exist for the identification of an operator error and -

correction of this error before any significant increase in the consequences of this event would occur."- The operator errcr referred to in this statement is a failure to isolate the _

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Lietdown line within 10 minutes of reaching saturation as assumed in the analysis. - A:

failure of the operator which results in the letdown line never being isolated is not considered credible due to operator training, the multiple procedure steps that require-letdown line isolation, and the inclusion of RBIC actuation (which closes the letdown -

line containment isolation valves) in the EOP " Rules" that are located in EOP-13.

These " Rules" are always applicable whenever the EOPs or abnormal procedures

(APs) are in effect, and the operator is required to memorize them. Rule 1, which

contains manual RBIC actuation on a loss of adequate subcooling, is emphasized during-operator training.

In the unlikely event that the operator would fail to isolate letdown within 10 minutes of the RCS reaching saturation, as assumed in the analysis, leakage will continue until letdown is isolated with a corresponding increase in offsite doses. However, the offsite ,

, dose consequences are approximately 1% or less of the 10 CFR 100 limits for a release that continues for 1170 seconds (19.5 minutes). The release could continue for significantly longer and still remain well below the 10 CFR 100 limits, providing time for the operator to recognize that an RCS leak exists and take action to isolate the letdown line as directed by procedures.

Numerous indications and alarms are available to alert the operator of a loss of reactor

- coolant in the event of the postulated letdown line break. These include decreasing ,

RCS pressure eventually resulting in a reactor trip, decreasing pressurizer level, decreasing makeup tank level, increased makeup flow, and a loss of adequate nsubcooling. -The operator would have indications that the RCS leakage is due to a

!-  : break in the letdown line via the area radiation monitors in the auxiliary building, the L effluent radiation monitors in the auxiliary building ventilation stack, auxiliary building sump alarms,' and visual and audible indications of a break in the letdown line (the auxiliary building operator would normally be in the area).

E.-  : If operator action is not taken, will automatic isolation be reached? And, why L was loss of pressure isolation setpoint not revised instead of the compensatory measures proposed with the manual operator action?

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_ 13F1197-26' Page 11 'j FPC Response: -

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The small increases in mass release and consequences for the changes in the letdown -

line break transient do not warraniconsideration of a change in the ES actuation'- ,

'setpoint as the transient does not challenge core ~ cooling.

As discussed in the response to question 9.D, numerous indications and alarms exist to  :

alert the operator to the postulated break in the letdown line, including a reactor trip j and a loss of adequate subcooling, which require implementation of the EOPs. Due to the multiple procedure steps that require letdown line isolation and extensive operator training, a failure of the operator to isolate a break in the letdown line is not considered-

' - credible.- Also, since the letdown line_ is equipped with multiple containment isolation valves, it can be isolated in the event of a single failure. 1

- EOP-13, Rule I would be implemented immediately if SCM was lost. Rule 1 calls for

' manual RBIC actuation which will isolate letdown. Loss of subcooling margin is the entry condition for FOP-3. EOP-03, Step 2.1 directs tripping RCPs and actuating HPI and RBIC. ~ EOP-3, Step 3.4 also directs the isolation of letdown, in the last bullet.

Step 3.9 is an "if at any_ time" step to ensure ES systems have or should have actuated.

Although these steps do not direct initiation of RBIC, it would be a prompt for the crew to remember they should have actuated RBIC manually for a loss of SCM. EOP-3, Step 3.16 ensures control complex Chillers are running. Step 3.17 then isolates

. potential RCS leak paths which includes MUV-38,-39 and 498, the letdown cooler inlet valves. This action would also isolate le!down. Even then, ifletdown was not isolated, Step 3.18 is an "if at any time" step to transition to EOP-07, " Inadequate-Core Cooling," for the RCS at greater than 20 degrees superheat. Step 3.2 of EOP-07 directs isolation of letdown.

FPC appreciates the opportunity to provide the above additional information. Please contact -

iMr, Dave Kunsemiller, Manager, Nuclear Licensing at (352) 563-4566 if you have any

_ questions regarding this response.

1 Sincerely, e

J.J. Holden l Director, Site Nuclear Operations JJII/TWC-Attachment t cc: l Regional' Administrator, Region II -

l Senior Resident Inspector NRR Project Manager -

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U.S. Nuclear Regulatory Commission 3F1197 26 Attachinent List of Regulatory Commitments The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager. Nucleat Licensing of any questions regarumg this document or any associated regulatory commitments.

ID Number Commitment Duc Date 3Fil97 26 The post-modification functional testing NovenSocr 10,1997 associa'ed with FPC Restart issue R-12 will assure overall balancing and proper flow rates for the Control Complex Ventilation System. The configuration and model for the control room dose calculations will be described in a separate submittal e,ssociated with the control complex habitability envelope (CCllE),

currently scheduled to be sent to the NRC on November 10,1997.

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