ML20029B158

Annual Rept of Personnel Exposure,Reactor Vessel Surveillance Capsules,Challenges to Pressurizer Relief & Safety Valves & Facility Changes,Tests & Experiments. W/910222 Ltr

ML20029B158
Person / Time
Site:	Crystal River
Issue date:	12/31/1990
From:	Boldt G FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0291-06, 3F291-6, NUDOCS 9103060027
Download: ML20029B158 (26)

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• Power C 0 h P O H A 1 s O ev Crystat tiver Unit 3 taket No. 50 302 february 22, 1991 3F0291-06 U. S. fluclear Regulatory Commission Attn: Document Control Oesk Washington, D.C. 20555

Subject:

1990 Annual Reports

Dear Sir:

The following annual reports at e submitted for the year 1990,

1. Annual Report of Persoinel Exposure in accordance with 10 C'<R 20.407 and Technical Specification 6.9.1.5.(a).

2. Annual Report of Reactor Vessel Surveillance Capsules in accordance with Technical Specification 6.9.1.5.(b).

3. Annual Report of Challenges to Pressurizer Relief and Safety Valves in accordance with Technical Specification 6.9.1.5.(e).

4. An' ;ual Report of facility Changes, Tests, and Experiments in

.:;ordance with 10 CFR 50.59(b).

Sincerely, G. L. Boldt Vice President liuclear Production Attachments

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GLB/JWT xc: Regional Mministrator Senior Residei, 1r.:Mctor tiRR Project liaaager [l 910306002/ yolg31 f PDR ADOCK 05000302 R pDR A Florida Progress Company

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ANNUAL REPORT OF ,

WHOLE BODY EXPOSURE AT CRYSTAL RIVER UNIT 3 JN ACCORDANCE WITH 10 CFR 20.407 REPORTING PERIOD JANUARY 1, 1990 TO DECEMBER 31, 1990 If personnel monitoring was not required to be provided to any individual during the year, checking this box will constitute a negative report indicating that such personnel monitoring was not required. .

OTHERWISE, COMPLETE THE FOLLOWING TABLE:

Estimated Annual Whole Body Number of Individuals

,_bn0sure Ranaes' (Rems) in Each Range No Measurabl e Exposure. . . . . . . . . . . . . . . . . . . . . . . . 104 7 Less Than 0.100................................ 525 1 0.100 - 0.250..................... ............ . 332 '

O.250 0.500.................................. 266 0.500 - 0.750.................................. 111 0.750 1.000.................................. 89 i 1.000 2.000.................................. 113 2.000 - 3.000.................................. 5 3.000 4.000.................................. 0 4.000 - 5.000.................................. 0 5.000 - 6.000.................................. 0 6.000 - 7,000.................................. 0 7.000 - 8.000.................................. 0 ,

8.000 - 9.000.................................. 0 9.000 - 10.000................................. 0 10,000 - 11.000................................ O 11.000 12.000................................ O Greater than 12.000............................ O TOTAL NUMBER OF INDIVIDUALS REPOR1ED.. ....... 2488 The above information is submitted for the total number of individuals for whom personnel monitoring was: (check one)

Required under 10 CFR 20,202(a) or 10 CFR 34.'1(a) during the calendar year.

X Provided during the caler.dar year.

• Individual values exactly equal to the values separating .1xposure ranges shall be reported in the higher range.

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ANNUAL REPORT Of >

4 RADIATION EXPOSURES AT CRYSTAL RIVER UNIT 3  ;

18 ACCORDANCE WITH TECHNICAL SPECIFICATION 6.9.1.5fa) d REPORTING PERIOD 3

JANUARY 1, 1990-10 DECEMBER 31, 1990 i

1. Personnel-(>100 Mrem) Total Man Rem r Work & Job function Utility Station Contract Utility Station Contract .

Reactor Operations & Surveillance  ;

Maintenance & Construction 1 5 2- 1.146 3.465 0.635 Operations 0 0 0 0.000 0.009 0.000 Health Physics & Lab 0 2 0 0.000 1.845 0.162 Supervisor & Office Staff 1 0 0 0.146 0.198 0.360 Engineering Staff 0 0 0 0.000 0.039 0.007 Routine Plant Maintenance ,

Maintenance & Construction 112 137- 303 51.607 42.340 173.256 ,

Operations 0 39 0 0.000 9.329 0.057 Health Physics & Lab 0 45 121 0.080 27.656 71.204 Supervisor & Office Staff 1 5 21 0.626 2.734 6.652 Engineering Staff. 5 1 33 3.221 0.420 16.743  !

Inservico inspection Maintenance & Construction 1 0 17 1.659 0.291 6.242 Operations 0 2 0 0.000 1.619 0.000 Health Physics & Lab -

0 0 0 0,000 0.059 0.354 4

Supervisor'& Office Staff 1 0 4 0.150 0.071 1.971 +

! Engineering Staff 0 0 6 0.075 0.000 3.057 i 6

Special,P1 ant Maintenance  ;

Maintenance & Construction 0 0 0 0.000 0.000 0.000 t Operations 0 0 0 0.000 0.000 0.000 Health Physice & Lab 0 0 0 0.000 0.000 0.000 Supervisor. & Office Staff 0 .0 0 0.000 0.000 0.000 Engineering Staff 0 0 0 0.000 0.000 0.000 .

Waste Processing _ _

Maintenance & Construction 0 0 1 0.000 1.114 0.801 Operations 0 0 0 0.000 0.000 0.000 Health Physics & Lab 0 0 2 0.000 1.550 1.172  ;

Supervisor &~0ffice Staff 0 0 0 0.000 0.351 0.000 Engineering Staff 0 0 0 0.000 0.000 0.000 Refueling-Maintenance & Construction 3 3 30 - 4 2.530 26.995

~0perations 0 0 0 - 30 0.036 0.000 Health Physics & Lab 0 0 0 .000 0.221 1.991 Supervisor & Office Staff 0 0 3 0.145 0.071 1.115 Engineering Staff 0 0 4 0.102 0,000 5.581 Totals-L Maintenance & Construction 116 145 353 57.146 49J41 207.929 Operations: - 0 41 0 0.000 10.994 0.057 Health Physics & Lab 0 47 123 0.080 31.331 74.883

, Supervisor & Office Staff 3 5 28 1.068 3.425 10.097

Engineering Staff 6 1 43 3.398 0.459 25.388 Grand Totals 125 239 547 61.692 95.950 318.354 Page 2 of 2 l

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5 ANNUAL REPORT Of REACTOR VESSEL MATERIAL SURVEILLAN',C CAPSULES IN ACCORDAt(CE WITH TECHNICAL SPECiflCATION 6.9.1.5(bJ REPORTING PERIOD JANUARY 1, 1990 TO DECEMBER 31, 1990 LESSIES CURRENTLY INSTALLED IN THE REACTOR Holder Position in Capsules lyht lloider Tube Installed ZY Top OC3 f Bottom CR3 A XW Top CR3 E Bottom TH12 D YX Top A2 Dottom A4 YZ Top TM12 LG1 Bottom TMl? LG2 WZ Top OC3 E Bottom CR3 LG2 WX Top OC3-C Bottom TMll F INSERTION / WITHDRAWAL

SUMMARY

END Of SEVENTH Cl(LE Holder Position In Capsule Capsule 11the Holder Tube Withdraw Insert XW Bottom TMil-B TM12 D

• YX Top TH12 0

• A2 YX Bottom OC3 D' A4 i Top WZ OC2 B OC3 E

• Capsulo TH12 D was moved from the top of the YX Holder Tube to the bottom of the XW Holder Tube.

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QRh[R$.tllP Of CAPSIlg.$ e florida Power Coroo.tation CR3 A CR3 E I

Duke Power Comngay 0C3 C OC3 E OC3 F CPU Nuclear Corngration l

TMil-f  :

TM12 D .

B&W Owners Orggp A2 A4 CR3-LG2 TM12 LG1 >

TM12 LG2 l' Page 2 of 2

ANNUAL REPORT Of CHALLENGES TO PRESSURIZER RELIEF AND CODE SAFETY VALVES AT CRYSTAL RIVER UNIT 3 IN ACCORDANCE WITH TECHNICAL SPECIFICATION 6.9.1.5 fe)

REPORTING PERIOD JANUARY 1, 1990 TO DECEMBER 31, 1990 Crystal River Unit 3 did not have any automatic Reactor Protection System trips in 1990. All the plant shutdowns were controlled. Therefore, there are no indications of challenges to the code safety valves, RCV 8 and RCV 9. RCV 8 was replaced in 1990.

The power operated relief valve (PORV), RCV-10, was only operated (challenged) as part of Plant Operating Procedure OP 202, Plant Startup, on the following dates:

1. 02 17 90 -

2. 06-18 90

3. 10 18 90

4. 10 24 90

5. 12 17-90 Page 1 of 1

ANNVAL RtPORT OF FAClllTY CHANGES, TEST, AND EXPERIMENTS AT CRYSTAL RIVER UNii 3 .

i lif ACCORDANCE W1J1LlQ_ CFR 50.59f b), i REPORTING PERIOD JANUARY l DECEMBER 31, 1990 In the attached report, each number refers to the Safety Evaluation questions listed below:

1. Is the probability of occurrence or the consequences of an accident or malfunction eva.luated in the FSAR increased?

YES , NO

2. Is the possibility for an accident or malfunction of a different type than any evaluated in the FSAR created?

YES , N0

3. Is the margin of safety, as defined in the bases for any Technical Specification reduced?

YES , NO .

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i 10 CFR 50.59 EVALVATIONS PERf0RMED  :

ON PLANT MODIFICATIONS IN ACCORDANCE WITH MAR PROCEDURES TACIllTY MODiflCM10N 4 MAR 81 10 19 ADDITION Of HIGH ENERGY LINE BREAK RESTRAINTS ON LINES TO EFP TURBINE SAFETY EVAtVATION

1. No, the 6 inch Main Steam Lines to the Emergency feedwater Pump (EfWP) are being analyzed as high energy lines in order to eliminate condensation. During testing, the turbine driven EFWP was tripping due to a large amount of condensation existing in the lines when MSV 55 and MSV 56 were closed. By opening MSV 55 and MSV 56 and analyzing the Main Steam Lines as high energy, the condensation in the lines will be eliminated. Thus, the probability of a malfunction of the turbine driven EFWP will not be increased.

2. No, since the 6 inch Main Steam Lines to the turbine driven EFWP are now high energy, break locations have been postulated. The possibility of a break occurring on these lines and a Jrt striking equipment important to safety was not previously evaluated in the FSAR, Lowever it was reported in LER 881 068.

This design package provides resolution to the LER and is consistent with previous safety anhlysis methodology.

3. No, the margin of safety is not reduced by this modification since pipe restraints are being designed to restrain pipe movement, and equipment that could be hit by a jet is being relocated.

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FACill1Y MODirirA110N MAR 63 11-14 01 l' ADDITION Of NUCLEAR INSTRUMENTATION SYSTEM SAFETY EVALVATION

1. No, the additional Nuclear Instrumentation System will in no way affect present i flux monitoring systems. The added system provides redundant, safety related, j

, backup, neutron flux monitoring in the Control Room. The added system is  !

passive and incapable of any control actions.

2. No, the added system is totally passive for redundant, safety related monitoring only and is incapable of any control action.

3. No, the added system does not affect any technicni specification and is totally separated from all present flux monitoring systems.

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IACJ111V N001LifAUD!i l

MAR 85 04 16 02 CHANGE COOLING WATER SUPPLY FOR CHEMICAL ADDITION HEAT EXCHANGERS 5, 6. AND 8 l SAFETY EVALUATigj

1. No, the consequences of an accident have been decreased. This MAR involves

changing the cooling water medium for CAHE 5, 6, and 8 fro.n Chilled Water to SW.

By removing the CH/ PASS sample interface the chances of contamination spreading outside the RCA, should a tube leak occur in CAHE 5, 6, and 8 are decreased.

2. No, new accident conditicas are not created. Thfs MAR involves changing the PASS sample cooling water for CAHE 5, 6, and 8. No change in the basic PASS system function is involved. The cooling water system is tming changed to address a potential tube leak contamination concern.

3. No, installation of this MAR will increase margins of safety. The possibility of contaminatina the Control Complex Ventilation, Switchgear Ventilation, and Penetration Cooling systems will be decreased, should a tube leak occur in CAHE 5, 6, or 8 with the cooling water changed from chilled water to SW.

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1 fAUllTY MODiflCATION J MAR 87 08 03 01 LETDOWN COOLER 3C ADDITION o

. SAFETY EVALVAT108

1. No, the modification made by this MAR is to add a third Letdown Cooler in '

parallel with existing Letdown Coolers 3A and 38. The design of this third

Letdown Cooler and its associated piping, valves, and components will essentially duplicate the arrangement of the existing Letdown Coolers with a few 1 minor variations. The Letdown Coolers are not required for safe shutdown although the letdown piping forms part of the reactor coolant pressure boundary.

This piping is, therefore, designated N1/S1 up to and including the new inside containment isolation valve which duplicates the original design. This new containment isolation valve receives an ES signal to close and is j environmentally qualified.

1 The Nuclear Services Closed Cooling Water System (SW) has been hydraulically analyzed to ensure that the addition of this new Letdown Cooler will not reduce the capability of the SW System to perform its normal cooling function. The

Letdown Coolers are isolated from the SW System after an accident by the i containment isolation valves which receive an ES signal. TF wefore, there is no flow to, nor is flow required to, the Letdown Cooler-post accident.- The portion of the SW System which is effected by this modification is not required -

for safe shutdown, FSAR sections evaluated with regard to these changes are 5.3, 9.1, 9.5, and Chapter 14. The accident evaluated in the FSAR Chapter 14 Safety Analysis, are not affected by this modification. Therefore, this modification will not i increase the probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the final Safety Analysis Report.

! 2. No, this modification does not create any new situation that would cause the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR. The modifications duplicate the design of the existing system which has already been considered in the Chapter 14 Safety Analysis (See item 1 above). No new unresolved safety questions are involved as a result of this modifica? an.

3. . No, the margin of safety as defined in the basis for the Technical Specifications is not reduced. The SW System will have sufficient cooling capacity available for continued operation of safety related equipment during normal and accident conditions as required by Technical Specification Bases Section 3/4.7.3. .The Technical Specification Bases for MU System is not

effected by this modification. The installation of the new letdown containment

isolation valve will ensure that the requirements of Technical Specification i Bases Paragraph 3/4.6.3 are met.

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[AUL11Y. MODiflCATION MAR 88 02 18 01 i EMERGENCY DIESEL GENERA 10R CRANKCASE PRESSURE

SWITCH REORIENTATION AND CONTROL LOGIC M00lflCAT10N SAFETY EVALVA110!j

1. No, this modification will re orient the crankcase pressurc switches (DL 13 PS through DL 18 pS) from a horizontal mounting position to a vertical mounting position. The existing engine control logic will be modified by adding the shutdown capability of the engines on high crankcase pressure in Testing Modes only. At the present time, the pressure switches wl'1 only initiate an alarm on high crankcase pressure (when any of the three pressure switches per engine actuate). No shutdown capability of the engines on high crankcase pressure exists. This modification will allow the shutdown of tne engines on high )

crankcase pressure (when any two of the three pressure switches actuate) during the testing medes of the engines only. The shutdown capability of the engines will be administrat_ively controlled via a key switch. The key switch will have two positions:

capability enabl (ed.1) TheNormal-Shutdown test switch will be keptcapability in thebypassed Normal position and (2)atTest all Shutd times except during the testing modes of the engines. The shutdown capability l 1s being added to the engines to prevent Paine damage during testing, but '

bypassed at all other times so that the ew bres will function the same during  !

Emergency and other Off Normal condition. l

2. No, the shutdown capability of the engines will be administratively controlled I and only enabled during the testing of the engines. The shutdown capability (to l.

l prevent engine damage) can be bypassed at any time (switched to Normal position) and allow the engines to operate the same as presently designed for Emergency and Off Normal conditions.

3. No, the operability requirements of the engines will not be affected by this t

modification. The engines will still operate as presently designed during any i Normal, Emergency, and Off-Normal conditions. The modification to the control logic will prevent any engine damage due to high crankcase pressure during the

. testing of the engines.

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l l [ACILITY MODIFICATIO.N .

MAR 88 07 06 01 GAG CHV 56, 57, 58, 59 IN POSITION t

. SAFETY EVALVATION i 1. No, these gags will assure that the chilled water valves will remain functional (CHV 56 & $8 open, CHV-57, 59 throttled) on loss of air to the actuators, thus helping to maintain the Control Room HVAC air temperature at its selected level.

The valves will continue to perform their design function including the ability to throttle, i

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2. No, because the valves are being modified to prevent closure or going to full i bypass posit sn on loss of air. This MAR will improve control room habitability '

, should it be necessary to keep valves in position if the air supply to the operators is lost. '

3. No, the modified valves will continue to perform their intended function,  !

including throttling. One chiller is capable of supplying both AHHE 5A & 50.

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flDLULh001LI.CAIJD!f

) MAR'38 08 01 01

, REACTOR BUILDING ENGINEERED SAFEGUARDS COOLING FAN SELECTABILITY

$AFETY EVALUATION

1. No, safety analyses for a LOCA (FSAR Section 14.2.2.$.3) have conservatively ,

assumed one R8 fan and, for an SLB (FSAR Section 14.2.2.1.5), no R53 cooling for I

maintaining the integrity of the Renctor Building with respect to pressure.

G/C, Inc. Calculation CR3.50ll.078-1, dated April 14, 1983, identifies that only ,

l one RB fan is needed in the event of an accident to maintain the RB temperature within design limits. Therefore, the probability of an occurrence or 4

consequence of an accident is not increased.

2. No, FSAR Sections 14.2.2.3.3 and 14.2.2.1.5 and G/C, Inc. Calculation CR-

' 3.5011.078 1, dated April 14, 1983, have identified that only one RB fan is i needed to maintain the integrity of the Reactor Building. Therefore, the possibility for an accident or malfunction of a different type is not created.

3. No, this modification does not impact any Technical Specifications requirements.

Therefore, margins of safety are not reduced.

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[EILITY MODIFICAT10B a MAR 90 03 20 01

! AUTOMATIC CONTAINMENT ISOLATION SYSTEM j INSTRUMENTATION ACCURACY .'iPROVEMENT SAFETY EVALVATION

l. No, the function of the Automatic Containment Isolation System (ACIS) as described in FSAR Section 9.4.2.7 is not changed. This modification to improve instrumentation accuracy does not change actuation circuits or valve operation.

Signal Isolation is provided from the remote shutdown system, therefore this equi;; ment important to safety is not affected.

2. No, the ACIS is not identified in the accident analysis, Chapter 14 of the FSAR.

The setp. int is changed because instrument accuracy is improved; the setpoint margin is maintained. The modification does not degrade the performance or increase challenges to any safety system.

3. No, the margin of safety provided by the setroint in Tech Spec 4.5.2.e is maintained.

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4 BCMtLE0JIFICATION ,

MAR 90 14 09 01 i INCREASE DHV 44 REllEf SETPOINT j grETY EVALVATION

1. No, as a result of the increase in the set pressure of DHV 44, affected piping and associated valves may be exposed to hi P r pressures under static conditions. At the new set pressure of 360 p ig, it has been conservatively

' assumed that line pressure may reach 40 ps 9 , which is the pressure corresponding to full relief. T'e potent al for Ligher static pressure in the affected portion of piping has be.n evaluated ani has Men found acceptable.

No overstress of piping or components due to the change in set pressure wO1 occur under .e ' tic conditions. Therefore, t% Ivobability of a line break, system malfune ion, or equipment failure due to piping / component overstress, or any other possible event or malfunction which could occur as a result of increasing the static pressure will not be increased. During heatup and cooldown, the DH suction side pressure will be administratively limited to 345

. asig. Limited system suction side operation at 345 psig has been evaluated and las been found acceptable with respect to code allowances for combined stress addressed in B31.1 paragraph 102.2.4. During normal system operation, ACIS will be available to protect the DH system from overpressurization. Since the pressure boundary integrity of the DH suction piping and components will be maintained under all conditions, the system's capability to perform its required safety functions will not be affected. Additionally, none of the individual l_

safety functions performed by system components, such as containment isolation, will be affected. Thus, the probability of occurrence or the consequence of an accident or equipment malfunctiori as evaluated in the FSAR will not be increased.

l 2. No, the setpoint change to be made will be made per an existing, approved plant procedure (MP 119). The only system / component parameter affected by the change in set pressure is the maximum pressure which may be present in the system piping under static conditions. Based on conservative analyses performed by r Gilbert / Commonwealth, the piping and associated component will not be l overstressed by the increase in the allowable static pressure. No changes to system operating procedures will be recuired to reflect the change in relief valve set pressure. Additionally, no cTanges will be made to the function of the system or the individual components. Therefore, since the integrity of the system will be maintained and no operating changes will be made as a result of the relief valve setpoint change, which could result in an abnormal operating condition, the possibility of a new accident or equipment malfunction will not be increased.

l 3. No, valve DHV 44 will be removed from its location in the DH system so that the l set pressure may be adjusted and testing performed per MP 119. OHV-44 is located on the dropline from the RCS and is common to both loops of the DH system. This presents a problem ir that a minimum of one loop of the DH system is required to be operable in all modes by the Technical Specification.s listed above. For this reason, the change in setpoint on DHV-44 will be accomplished while the Unit is defueled during Refuel 7.

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FAClt 1TY E9911j[El%j MAR 90 04 09 01 INCREASE DHV-44 RIllEf SCTI'0lNT

$Al[LVELUATION (Continued) l'i the defueled configuration, no limiting conditions for operation as reiquired by the Technical Specifications are applicable. Following cer.pletion of the set pressure change, the relief valve will be tested to verify the accuracy of the new set pressure. The YalVe Will then be reinstaIM 00d will be Visuolly i checked for leaks when the system / loop is placed back in service. As stated in the response to questions 1 and 2 above, the integrity of the DH system will not be affected by the setpoint change. The OH system and affected components will not be adversely affected by the change in set pressure and will be fell't cap 4 hie of performing their required functions. Therefore, based on performance of field work in 6 no mode situation and full restoration of system operability following completion of work, the margin of safety as established by the Technical Specifications will be maintained. >

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FACillTY MODiflCATION i MAR 90 04 11-01 EDG FVEL TANKS CROSS-CONNECTION SAFETY ;yALVATIQN

1. No, a malfunction of components to be installed in the cross connection between fuel oil Day Tanks A and B will not affect the safety of the Emergency Diesel Generator System as described in the FSAR Soc!. ions 1.4.3, 8.24 3.1, 9.8. 7.5 and l 9.8.8, and detailed in item 3 of the Design triput Record. l l

2. No, installation of a cross connection between fuel oil Day Tanks A and B will not create an accident not previously addressed in FSAR Sections 5.4.4, 8.2.3, r-d 14.2.

3. No, cross connecting fuel oil Day Tanks A and B will not adversely affect the Emergency Diesel Generator System addressed in the Electrical- Power Systems

( ' tion of the Technical Specifications at 3.8.1.1 and 3.8.1.2.

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10 CFR 50.59 EVALUATIONS PERFORMED ON FSAR TEXT CHANGES IN ACCORDANCE WITH PROCEDURE N0D 11 FSAR SECTION 3,9 - CYCLE 8 SAFETY EVALUATION

1. No, it is concluded from the review of the Cycle 8 Reload Report (BAW-2102, Rev,

1) that Cycle 8 core thermal and kinetics properties. with respect to acceptable previous cycle values, da not adversely affect the ability of the Crystal River 3 plant to optrate safely during Cycle 8. Considering the previously accepted design basis c ad in the FSAR and subsequent cycles, the transient evaluation of Cycle 8 is bounded by previously accepted analyses. The initial conditions for the transients in Cycle 8 are bounded by the FSAR, with exception of the end of Cycle (E00) hioderator temperature coefficient. The E00 temperature coefficient will remain within the current FSAR hot full power limit for the steam line break and dropped rod analysis at a soluble boron concentration of 84 ppmB or highei. Additional analysis will be required to support Cycle 8 operation prior to reaching a soluble boron concentration of 84 ppm 8 or below. l It is concluded, from a comparison of the radiological doses presents in the FSAR to those calculated specifically for Cycle 8, that although some accident doses exceed the FSAR values, these increases are not significant. All doses are less that the 10 CFR 100 limits, and meet acceptance criteria.

2. No, the Cycle 8 reload fuel design is of the same type as found acceptable for prior cycles, consequently the probability of an accident or-malfunction of a different type than previo,isly evaluated is not created.

3. No, the Cycle 8 reload parameters have been reviewed with respect to applicable Technical Specification Bases and it is concluded that the margin of safety for Cycle 8 is not reduced.

Page 13 of 20

FSAR SECTIONS 5.6.4.1. 2. and 3: 14.2.2.5.9: and Table 14 46 - R.B PRESSURE SAFETY EVALVATION

1. No, the probability of an accident or malfunction previously evaluated in the FSAR occurring will not be increased since this FSAR change does not describe a modification of any equipment or a change in normal plant operating conditions.

The consequences of an accident or malfunction previously evaluated in the FSAR will not be increased since the radiological consequences of a Design Basis LOCA and a Maximun Hypothetical Accident have been analyzed at a containment leak rate corresponding to the design pressure and temperature of the Reactor Building. Increasing the test pressure to a value still below design pressure ensures that the leak rate used in the analysis still envelopes the leak rate at worst case post-accident containment pressure.

2. No, this FSAR cha-qe will not create an accident or malfunction different than that evaluated in the FSAR since no equipment has been modified and no normal plant operating conditions have been changed.

3. No, the margin of -safety, as defined in the bases of the Technical Specifications, will not be reduced since the new test pressure is still below design pressure and, therefore, the containment leak rate used in the accident analyses (based on design pressure) is still enveloping.

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FSAR TABLE 9 MU SYSTEM PERFORMANCE DATA l

i SAFETY EVALVATION

1. No, this change is editorial in nature to provide clarification of RCP seal flow data. No modification to the plant is involved.

2. No, this change is editorial in nature for clarification. No plant tuodification is involved.

3. No, this change is editorial in nature for clarification. No plant modification is involved.

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j FSAR SECTIONS 1.2.6 AND 1.9 SAFETY EVALUATION

l. No, change to FSAR corrects descriptions of CR 3 interface with 230 KV l switchyard and other CR units. Change is nomenclature for switchyard supply lines and corrects interface in Chapter 1 with CR 1 and 2. Changes to Unit I and 2 interface were previously addressed in FSAR submittal for offsite power PAR 89-08-11-03.

2. No, configuration change was done and no revalidation of any scenario previously identified in Chapter 14 is required.

3. No, changes made to FSAR add consistency to descriptions of CR-3 interfaces.

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FSAR SECTION 1.4 - GENERAL DESIGN CRITERIA SAFETY EVALVAlj.0.3 l

1. No, these changes are editorial and do not affect the probability of occurrence of an accident or malfunction of equipment previously evaluated in the FSAR.

2. No, these changes clarify NRC 10 CFR 50, Appendix A requirements and do not deal with FSAR accident consequences or equipment evaluated in the FSAR.

3. No, these changes do not affect the Technical Specifications.

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FSAR TABLE 7

SUMMARY

OF SHVT00 WILE 0VIPMENT SAFETY EVALVATION

1. No, this is an editorial change to correct the DCP-1A equipment title. Table 7-10 identifies DCP 1A as "RB Fan Assembly 3A" versus "DH Closed Cycle Cooling W9ter Pump "

2. No, this is at. editorial change to correct the equipment title of DCP 1A.

3. No, this is an editoriel change to correct the equipment title of DCP-1A.

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FSAR SECTION 9.8 - FIRE PROTECTION SAFETY EVAltJATION

1. No, the changes to the Fire Protection Plan are conservative in nature and when addressed by the FSAR are compatible with that document. They encompass programmatic and administrative changes required by regulatory documents which enhance the Fire Protection Program. A review of the FSAR indicates that the chat.ges do not affect the performance of any equipment of alter the evaluation previously made in the FSAR regarding the probability of occurrence or consequences of an accident or malfunction of equipment.

2. No, the changes to the Fire Protection Plan do not introduce any unevaluated programmatic or administrative conditions which could affect the performance of any equipment previously evaluated or cause an accident of a different type than previously evaluated by the FSAR.

3. The Fire Protection Plan is not Technical Specification related and no margin of safety issues are affected.

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SECTION 4.2.1.1 - RCS SYSTEM DEK.RIPTION SAFETY EVALVATION

1. No, this revision to the FSAR will correct one item, delete one item, and augment the information currently contained in FSAR Section 4.2.1.1. The following changes are being made to the FSAR section:

1. the number of letdown cooler outlet valves is being changed from 2 to 3,

2. the RCP seal injection valve is being deleted since it no longer receives an ES signal to close, and

3. valve tag numbers are being assigned to the valves which were previously identified only by type of service.

2. No, this revision will have no effect on the configuration or operation of any plant system, and, therefore, the possibility for an accident or malfunction of a different type than previously evaluated in the FSAR is not created.

3. No, these changes to the FSAR have no effect on the margin of safety.

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