ML20010F890

Forwards Response to Request Re Section 3 of Safety Evaluation for Environ Qualification of safety-related Electrical Equipment Transmitted in NRC 810521 Ltr.Response to Section 4 of Evaluation Will Be Submitted by 811008

ML20010F890
Person / Time
Site:	Point Beach
Issue date:	09/11/1981
From:	Burstein S WISCONSIN ELECTRIC POWER CO.
To:	Clark R, Harold Denton Office of Nuclear Reactor Regulation
References
IEB-79-01B, IEB-79-1B, NUDOCS 8109150080
Download: ML20010F890 (20)

Text

_

l WfSCOnSin Electnc eaare couvar 231 W. MICHICAN P.O. BOX 2046. MitWAUKEE, WI S3201 September 11, 1981 N co Mr. Harold R. Denton, Director 8 b' Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION

( / I NL 1/ j f f Washington, D. C. 20555 m

[F g_.

- i Attent!on: Mr. Robert A. Clark, Chief F. 3EP 141981s Mgp- J -/

Operating Reacters Branch 3

% ~'

Gentlemen: N

'b s DOCKET NOS. 50-266 AND 50-301 RESPONSE TO SAFETY EVALUATION REPORT FOR ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED EQUIPMENT POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 On May 21,1981, Mr. Robert A. Clark of your Staff transmitted to Wisconsin Electric Power Company the Safety Evaluation Report (SER) for the Environmental Qualification of Safety-Related Electrical Equipment at Point Beach Nuclear Plant, Units 1 and 2 (PBNP). You requested that we provide additional information identified in Sections 3 and 4 of the SER.

This letter transmits our response to your request.

Our responses to the requirements and/or questions list.ed in Section 3 of the SER are shown in Enclosure 1 to this letter. The responses

, are identified by the corresponding SER section number and title. Updated l " master list" equipment sheets, the corresponding component eabation

, worksheets, and related notes are provided in Enclosure 2. These updates i resulted from a revision to the PBNP Emergency Operating Procedures dated l July 15,1981.

l l Our remaining component evaluation worksheets updated to address l the requirements of Section 4 of the SER will be provided in a supplemental letter by October 8,1981.

Our equipment qualification evaluations to date demonstrate that Point Beach Nuclear Plant, Units 1 and 2, can continue to operate until the requirements identified in the SER are fully resolved. We would be pleased to answer any further questions concerning this response.

Very truly yours, Sol Burstein Exe tive Vice President Enclosures nt Inspector l \@kg r n109150080 81'Ok11 h' l DR ADOCK 05000266 W PDR  ;

)

ENCLOSURE;1 i

j

Enclosure 1 3 Staff Evaluation: This section listed deficiencies identified by Mr. F. J.

Jablonski of IE (Region III) in his site inspections of April 21 and October 10, 1980 which " included a lack of plant equipment identification numbers and nameplate data that did not correspond to the CES "[ Component Evaluation Sheets ] . " The lack of identification numbers was identified only because valve air-operators at PBNP have an ID tag attached to and the associated solenoid valves, I/P transducers, and/or limit switches are not separately tagged. ' Since these devices are physically connected to the operator, this method of identification is jedged to be acceptable. The one RTD (component cooling heat exchanger outlet) which was not physically tagged was adequately identified by physical location as confirmed in Mr. Hayes' (IE Region III)

October 21, 1980 memorandum concerning inspection of installed systems at Point Beach Unit 1. The only components for which the nameplate data did not exactly correspond were Westinghouse motors used for residual heat removal and component cooling. As stated in Mr. Hayes' October 21, 1980 memoran:ium, the Westinghouse model numbers were correct, but the insulation type on the submittal was not stamped on the motor nameplate. The insulation type for these motors was obtained by written correspondence from Westinghouse, copies of which are located in Wisconsin Electric's central equipment qualification file.

In addition, a question was raised as to the separation criteria applied to the arrangement of pump and valve motors of the safety injection and containment spray systems of Units 1 and 2. This issue was previously addressed in an April 9,1981 letter from Mr. C. W. Fay of Wisconsin Electric to Mr. Harold R. Denton of the NRC Staff. Therefore, all potential deficiencies identified in this section are considered resolved.

3.1 Completeness of Safety-Related Equipment: The NRC Staff requested a listing of all systems both inside and outside potentially harsh environments required to " achieve or support: (1) emergency reactor shutdown, (2) containment isolation, (3) reactor core cooling, (4) containment heat removal, (5) core residual heat removal, and (6) prevention of significant release of radioactive material to the environment." Based on a detailed review of the PBNP FFDSAR, Technical Specifications, and Emergency Operating Procedures, the following table lists all the systems that are required to perform those functions at Point Beach Nuclear Plant, Units I and 2 in the event of a postulated Loss of Coolant Accident (LOCA) or High Energy Line Break (HELB) accident. The corresponding nomenclature identified by the DOR Guidelines, Appendix A, Typical Equipment / Functions Needed for Mitigation of a LOCA or MSLB Accident are also shown:

PBNP SYSTEMS REQUIRED TO MITIGATE LOCA OR HELB ACCIDENTS PBNP System DOR Nomenclature Comments Safety Injection Emergency Core Cooling; Includes High-Head Containment Fission Safety Injection and Product Removal Containment Spray i

PBNP System DOR Nomenclature Comments

~

auxiliat y Coolant Emergency Core Cooling; Includes Low-Head Component Cooling; Safety Injection, Residual Heat Removal Component Cooling, and RHR (Note 1)

Chemical and Volume Chemical and Volume Only Boric Acid Control Control Tanks and associated equipment are safety-related Reactor Control and Engineered Safeguards Only components required Protcction Actuation; Reactor to initiate reactor trip, Protection; Emergency emergency safeguards Shutdown actuation, and containment isolation are safety-related Reactor Coolant Reactor Coolant (Pres- Note I surizer Sprays, PORVs)

Main Feedwater Main Feedwater Shutdown Only instrumentation and Isolation; Reactor and associated components Protection are safety-related and potentially exposed to harsh environments Main Steam Steamline Isolation; Only instrumentation Engineered Safeguards and valve operators for Actuation; Reactor steam supply to AFW pump Protection turbines are safety-related and potentially exposed to harsh environments.

Containment Air Containment Heat Removal Only emergency fan Recirculation Cooling cooler units are safety-related Containment Isoiation Containment Isolation; Only Containment Isolation Containment Ventilation valves and associated equipment are safety-related Auxiliary Feedwater Auxiliary Feedwater All electrical components except condensate storage tank level instrument a.re located in mild environments only Electrical Emergency Power Only safeguards motor control centers are safety-related and potentially exposed to harsh environments PBNP System DOR Nomenclature Comments Service Water Service Water All safety-related electrical components located in mild environments F.ailation Monitoring Radiation Monitoring; Only components required Containment Radiation to initiate containment Monitoring purge isolation are important to safety and these are located in mild environments only.

Note 1: The design and licensing basis of Point Beach Nuclear Plant, Units I and 2, is to maintain the ability to achieve and maintain a hot shutdown condition following any design basis accident.

Equipment required only to achieve a cold shutdown condition (e.g. , PORVs or the RHR system) is not safety-related.

Only those systems required for mitigation of a LOCA or HELB accident based on the PBNP Final Facility Description and Safety Analysis Report (FFDSAR), Technical Specifications, and Emergency Operating Procedures and whose components are potentially exposed to a harsh environment from those accidents were listed in our responses to IE Bulletin 79-01B. The environmental qualification of components required for Post-Accident Sampling and Monitoring, Radiation Monitoring (including containment), and Safety-Related Display Instrumentation are being addressed as part of evaluations and modifications being undertaken to meet the intent of the TMI Action

'; Plan (NUREG-0737). These evaluations and modifications will follow the schedule committed to in our numerous responses to the Staff related to the TMI Lessons Learned. The Heating, Ventilating and Air Conditioning (HVAC) systems for the control room and areas containing safety equipment are not located in potentially harsh environments and were, therefore, not identified in our system: list. Venting is used as the containment combustible gas control system at Point Beach Paclear Plant. This addresses all of the typical systems listed in Appendix A to tla DOR Guidelines.

The Waff also requested us to provide "a complete list of instrumentation m;ntioned in the LOCA and HELB emergency procedures." The following is a list of that instrumentation by PBNP tag number and function with justification provide <1 for not environmentally qualifying the instruments which are not safety-related:

Function PBNP Tag No. Comments RCS Temperature, l&2-TE450A&B RCS Loop RTDs are being qualified including l&2-TE451A&B rather than the bypass manifold RTDs THOT, TCOLD, because the loop RTDs can be used during natural circulation conditions.

and T AVG The operator will rely on the qualified instrument for operator action following the accident. Therefore, failure of the unqualified instruments will not adversely affect LOCA or HELB accident mitigation.

Function PBNP Tag No. Comments RCS Pressure l&2-PT420 Pressurizer l&2-PT429, 430, 431, Pressure and 449 Pressurizer Level 1&2-LT426, 427, and 428 Radiation Monitors These instruments are being  ;

(Contai*nment, Air evaluated and upgraded per Ejector, Steam TMI Lessons Learned. They Generator Blowdown) are not required to mitigate a LOCA or HELB accident per the PBNP FFDSAR or Technical Specifications. Therefore, failure of these instruments will not adversely affect LOCA or HELB accident mitigation.

Core-Exit l&2-TEl-39 These instruments are being Thermocouples evaluated and upgraded per and Subcooling the TMI Lessons Learned.

They are not required to mitigate a LOCA or HELB accident per the PBNP FFDSAR or Technical Specifications. Therefore, failure of these instruments <

will not adversely affect LOCA or HELB accident mitigation.

Steam Generator l&2-LT460, 461, 462, Level and 463 l&2-LT470, 471, 472 and 473 Main Steam flow 1&2-FT464 and 465 (i.e. , S/G Steam l&2-FT474 and 475 Flow)

Main Feedwater l&2-FT466 and 467 These instruments are not -

flow required to mitigate a LUCA or HELB accident per the PBNP FFDSAR or Technical Specifications. Main feedwater is secured automatically on these accidents. Therefore, failure of these instruments will not adversely affect LOCA or HELB accident mitigation.

)

4 Function PBNP Tag No. Comments Main Steam 1&2-PT468, 469, and 482 Pr. essure (i.e. , 1&2-PT478, 479, and 483 S/G Pressure)

Containment l&2-PT945, 946, 947, 948 Pressure 949, and 950 Containment Sump 1&2-LC942A&B B Wate,r Level l&2-LC943A&B Containment These instruments are not Temperature, required to mitigate a LOCA Humidity, and or HELB accident per the Sump A Level PBNP FFDSAR or Technical Specifications. The operator is not required to take accident-mitigating action based on these instruments.

Therefore, failure of these instruments will not adversely affect LOCA or HELB accident mitigation.

Boric Acid 1&2-LT106, 172, 190 Storage Tank LT-102, 171, 189 Level Auxiliary Feed- All components for these water Flow components are located in mild environments.

Volume Control Tank The components of the Charging Level and Volume Control System and Charging Flow; their associated instrumentation Letdown Flow; are not required to mitigate LOCA Charging Pump Speed or HELB accidents per the PBNP FFDSAR or Technical Speci-fications. The operator takes no required accident-mitigating action based on this instrumen-tation . Therefore, failure of these instruments will not adversely affect LOCA or HELB accident mitigation.

Refueling Water All components for these instru-Storage Tank Level ments are located in mild environments only.

Condensate Storage LT-4025 and 4031 Tank Level L

function PBNP Tag No. Comments Diesel Generator All components of these Output Voltage, instruments are located in Frequency, and mild environments only.

KW Load Safety Injection l&2-PT924 and 925 Pump Flow i

Safety Injection 1&2-PT922 and 923 Pump Discharge

, Pressure Low Head SI l&2-FT626 and 928 (RHR) Flow Low Head SI 1&2-PT628 and 629 (RHR) Pump Discharge Pressure RHR Temperature l&2-TE627 and 630 Containment l&2-LT931 Spray Additive Tank Level Component Cooling 1&2-PT619 Flow Component Cooling l&2-TE621 Heat Exchanger Outlet Temperature Containment Spray This instrument is not Additive Flow required to mitigate a LOCA l

or HELB accident per the PBNP FFDSAR or Technical Specifications. The operator is directed by the EOP's to spray additive valve position indicatica and spray additive tank level, l which are being environmentally i qualified, to verify that all l

' NaOH has been added to the containment spray. Therefore, failure of this instrument will not adversely affect LOCA or HELB accident mitigation.

l l

1 Function PBNP TE.g No. Comments Reactor Power Level; These instruments are not Control Rod Position required to mitigate LOCA or )

4 Indication; HELB accidents per the PBNP Pressurizer Temperature; FFDSAR or Technical Speci-Waste Holdup fications. The operator is Tank Level not required to take accident-mitigating action based on these instruments. Therefore,

' failure of these instruments Vil not adversely affect LOCA or HELB accident mitigation.

Only those instruments that are required for mitigation of design-basis LOCA and HELB accidents per the PBNP FFDSAR, accident analyses, Technical Specifications, and Emergency Operating Procedures and are potentially exposed to harsh environments from those accidents are considered within the scope of IE Bulletin 79-01B. Those instruments in the above list which were listed without comments were included in the master list of equipment under the plant system to which they belong. In addition, the RCS Loop RTDs and Core-Exit Thermocouples were included in the updated master list and component evaluation worksheets which were submitted to the NRC Staff on January 30, 1981 in response to IE Bulletin 79-OlB, Supplement No. 3. Our detailed evaluation of the remaining instruments in the above list leads to the conclusion that their failure will not affect the performance of any safety system or mislead the operator in such a manner as to adversely affect LOCA or HELB accident mitigation.

Wisconsin Electric is activei; pursuing the evaluation and upgrading of additional instruments in respo::se to the TMI Action Plan (i.e., NUREG-0737),

NUREG-0696, and Regulatory Gutie 1.97. The schedule for this effort, however, is consistent with Wiscol. sin Electric's commitments related to the above documents which were previously provided to the Staff.

3.2 Service Conditions: In its review of the service conditions inside containment, the Staff assumed that the Main Steam Line Break (MSLB) conditions are enveloped by the large-break LOCA conditions. The Staff required us to verify the assumption that PBNP is equipped wth an automatic containment spray system which satisfies the single-failure criterion. The PBNP FFDSAR, Section 6.4, Containment Spray System, Subsection 6.4.3, Design Evaluation, states "A single failure analysis has been made on all active components of the system to show that the failure of any single active component will not prevent fulfilling the design function. This analysis is summarized in Table 6.4-4." This statement and an additional design review verify that the Containment Spray System is initiated automatically and satisfies the single failure criterion of 10 CFR 50, Appendix A.

1 Therefore, the Containment Spray Systems at Point Beach Nuclear Plant, Units 1 and 2, satisfy the requirements of Section 4.2.1 of the DOR Guidelines.

L l

3.3 Temperature, Pressure, and Humidity Conditions Inside Containment:

The NRC Staff has now concluded, contrary to the requirements of the i

DOR Guidelines, Section 4.2.1, that the containment temperature profile for equipment qualification purposes should include an additional " margin to l

account for higher-than-average temperatures in the upper regions of the

' containment that can exist due to stratification, especially following a postulated MSLB. Use of the steam saturation temperature corresponding

] to the total building pressure (partial pressure of steam plus partial pressure 4

of air) versus time will provide an acceptable margin for either a postulated j

4 LOCA or MSLB. . ." The DOR Guidelines, Section 4.2.1, states that " equipment qualified for LOCA environment is considered qualified for a MSLB accident i environnient in plants with automatic spray systems not subject to disabling i single component failures." The Commission's Memorandum and Order i

CLI-80-21, dated May 23, 1980, and the Staff's Order for Modification of 1 License, dated October 24, 1980, Concerning Environmental Qualification of

]

Safety-Related Electrical Equipment at Point Beach Nuclear Plant, Units 1 and 2, both endorse the use of the DOR Guidelines or NUREG-0588, as i appropriate, for establishing the adequacy of environmental qualification.

It is Wisconsin Electric's position that the use of the containment temperaure profile calculated for a design-basis, large-break LOCA, which has been previously approved by the NRC Staff, is acceptable for equipment qualification inside containment. Therefore, the required temperature profile referenced in the component evaluation :vorksheets for equipment located inside containment j remains valid and will not be changed.

From a technical standpoint, the use of the saturation temperature corresponding

to total building pressure versus time is overly conservative except in two j cases. The first case is where equipment is located extremely high in l containment and stratification effects following a MSLB may cause temporary superheated conditions until the containment spray system starts at a maximum of one minute following the initiation of the accident. In this

case, the thermal lag of the safety-related equipment would prevent the critical internal temperatures from exceeding the equilibrium temperatures

, r ached during qualification to the conservatively-calculated LOCA temperature

, profile. Since a typical MSLB containment temperature profile (see NUREG-l 0458, Figure 1) for a Westinghouse PWR shows a temperature reduction to d

240 F or less almost immediately after spray initiation, the qualification to j the LOCA temperature profile provides adequate assurance of safety from i

an equipment qualification standpoint.

The second case is where equipment is located in the direct vicinity of a high-energy line with no physical barrier such as a wall or floor in between l and, therefore, could experience higher temperatures than calculated during a postulated accident. The superheated steam escaping from the break in this case may not mix with and/or be cooled by the containment air or surfaces before it reaches the safety-related equipment. A detailed review of equipment locations inside containment with respect to elevation I

and to vicinity of high-energy lines was conducted. The only safety-related equipment which falls into either of the above two cases is the solenoid

, valves and limit switches associated with the containment purge supply and

! exhaust valves, l&2-HV3213 and 3245. Those components are being replaced with components qualified to well above the maximum saturation temperature.

l L

1 l

I In the interim, continued operation is justified in spite of the higher temperature profile by the fact that purge supply and exhaust valves and their redundant backups outside containment are administratively " red-tagged" closed unless the plant is in a cold shutdown condition. Therefore, failure of these components could not adversely affect LOCA or HELB accident mitigation.

The Staff has also requested that the relative humidity associated with each service condition (i.e., temperature and pressure) profile be provided.

As stated on the component evaluation worksheet, a relative humidity of 100% was' assumed as the worst case. Temperature and pressure are the only parameters related to steam and moisture which are significant in qualification of equipment to LOCA or HELB accidents. In fact, the DOR Guidelines do not even mention humidity as a parameter to be considered in qualification except for equipment in confined spaces in areas where fluids are recirculated from inside containment to accomplish long-term core cooling following a LOCA (See DOR Guidelines, Section 4.3.2). Therefore, the identification of 100% relative humidity on the component evaluation worksheets is considered sufficient and the service condition profiles do not need to be changed.

]

The Staff also requested that a pressure, temperature, and humidity profile be provided "for all areas subject to a potential'HELB." In addition, room numbers or other applicable designation should be used to specify the area to which each profile applies. Temperature and pressure profiles were developed and submitted for all safety-related components potentially exposed to a HELB accident environment outside containment. The profiles were identified by component rather than room number because the profile is different for every location within the room. The pressure profile is a function of distance from the high-energy line due to expansion of the steam jet as a function of distance from any potential break location. There is no need to provide profiles for all areas. Humidity profiles are not required by the DOR Guidelines, as discussed in the above paragraph.

Therefore, our previous submittal is considered sufficient in this regard and no changes to existing profiles or additional profiles are provided.

3.4 Temperature, Pressure, and Humidity Conditions Outside Containment:

See the last paragraph of Section 3.3 response above.

3.5 Submergence

The Staff has requested additional information concerning the submergence of several components identified as having the potential for submergence. It was also requested that potential submergence of safety-related electrical equipment outside containment be addressed. The information requestz.3 will be provided with the updated component evaluation worksheets.

3.6 Chemical Spray: The Staff has requested the exact chemical concentration of boric acid and sodium hydroxide used in qualification testing and a i

further discussion of the effects of chemical spray. This information will i be provided with the updated component evaluation worksheets.

_g.

3.7 Aging

The NRC Staff has requested additional information to verify the degree of compliance with the aging requirements of the DOR Guidelines for safety-related electrical equipment potentially exposed to harsh accident environments. The requirements include the following actions:

1. Existing equipment must be analyzed to identify any "ruaterials which are known to be susceptible to signifitant degradation due to thermal and radiation aging." If the devicc contains such material, a qualified life "must be established on a case-by-case basis;"

2. Establish preventative maintenance, surveillance, and/or replacement schedules which take into account the specific aging characteristics of the installed equipment; and

3. Establish an ongoing program to review surveillance and maintenance records to assure that equipment exhibiting age-related degradation is maintained and/or replaced as necessary.

The first requirement to evaluate equipment for potential degradation due to thermal and radiation aging has been met by two methods. For equipment which was pre-aged prior to design-basis event testing, the pre-aging program was evaluated and compared to the plant-specific application to establish an expected minimum life including margin. For equipment which was not pre-aged during environmental qualification tests, the organic materials (i.e., those other than metals or ceramics) were identified and evaluated for potential degradation due to thermal or radiation aging in order to establish an expected minimum life including margin. Appendix C to the DOR Guidelines as well as other references were useu in these evaluations. Examples of both methods of establishing an expected lik. are shown in the notes for the updated component evaluation worksheets.

The second and third aging requirements from the DOR Guidelines concern maintenance, replacement, and surveillance programs. These programs already exist at Point Beach Nuclear Plant for the purpose of ensuring that safety-related equipment will be able to perform its design function throughout its installed lifetime. These programs and supporting administrative procedures, however, are being re-evaluated and will be upgraded, if necessary, before the environmental qualification deadline (presently June 30, 1982) to comply with the intent of the DOR Guidelines. Existence of these programs will be documented in Wisconsin Electric's central equipment qualificaticn file by the deadline and will be available for audit by the NRC Staff.

3.8 Radiation (Inside and outside Containment): The NRC Staff has found the required radiation doses used for equipment qualification inside and outside containment to be acceptable. Therefore, no further response is provided. The methodology employed in dose and dose rate calculations is maintained in our central equipment qualification file and is available at our general offices for NRC audit.

i t

- ~ - - ma .---- +--na- ..---.m--sa,aL-m.an p.,,---- 4 u e J J e-a J __,,, A. - -A- - L,mm-m---.--mA wm 1-.%--s._4- A5-- n - - - - -J- - -- - -ma---

E a

a i

4 a

)

F i

b

. t i

8 ENCLOSURE 2 4

l 1

i l

f l

{

i l

l b

ENCLOSURE 1 1 - FACILITY: Point Beach Nuclear Plant, Units 1 and 2 Page 1 of 15 MA TER LIST Rev. 5 DOCXFT NO.: 50-266 and 50-301 (CLASS IE ELECTRICAL EQUIPMENT REGUIRED TO FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS)

I. SYSTEM: Safety Injection COMPONENTS LOCATICN PLANT IDENTIFICATION INSIDE PRIMARY OUTSIDE PRIMARY NUMBER GENERIC NAME CONTAINMENT CONTAINMENT l-P14A&B Pump Motors X 2-P14A&B .

1-P15A&B 2-P15A&B Pump Motors X l-SI851A&B Valve Motor 2-SI851A&B Operators X l-SI871A&B Valve Motor 2-SI871A&B , Operators X l-SI860A, B, C, & D Valve Motor 2-SI860A, B, C, & D Operators X l-SI852A&B Vc.lve Motor 2-SI852A&B Operators X l-SI878A, B, C, & D Valve Motor 2-SI878A. B. C. & D Operators X 1-SI836A&B Electro-Pneumatic 2-SI836A&B Transducers X lamco " Snap-Lock" 32400X Limit Switches X l-PT9225923 Pressure

.  ?-PT922&923 Transmitters X l-FT924&925

?-FT924&925 Flow Transmitters X l-FT928

?-FT928 Flow Transmitters X l-LT931

?-LT931 Level Transmittore v l-LC942A&B, 943A&B

?-LC942A&B, 943A&B Level Switches X American Oil Co.

Amolith #2 Lubricant (Grease) X American Oil Co.

1molith #1EP Lubricant (Grease) X X l lobil Oil Co.

b28 Lubricant (Grease) X X American Oil Co.

Industrial #35 Lubricant (Oil) Y l 1merican 'il Co. Rykon l Industria 5 Lubricant (011)

X Okonite 5,000 Volt AC Power Cables L *

....-.....-...w-. . . . . . . . .c . . . .

FACILITY: Point Beach Nuclear Plant, Units 1 and 2 Page 15 of 15 50-266 and 50-301 MASTER LIST Rev. 5 DOCKET No.:

(CLASS IE ELECTRICAL EQUIPMENT RECUIRED TO FUNCTION UNDER POSTULATED ACCIDENT CONDITIONS)

XII. SYSTEM: Electrical COMPONENTS

. LOCATION PLANT IDENTIFICATION -

INSIDE PRIMARY OUTSIDE PRIMARY NUMBER GENERIC NAME CONTAINMENT CONTAINMENT l-B32 Motor Control Center X

?-B32 Motor Control Center X Kerite 600 V Power Cable X Rome 500 V Control Cable X e

9

-y- - . , - . - - . . , - , , . , , ._

, , , , .,- , , . - - , , . , , , , , . , , . , _ . . g -e- -v. --

FACILITY: Point Beach Nuclear Plant '

UNIT: 1 I.-20 Rev. 5 .

DOCKET: 50-266 S_YSTEM COMP _0NENT EVALUATION WORK SHEET

---~ *-~"~~ "

' ~~~ UUCUMUit ATIDFlill'[IIDiCES ENVIRONMENT (See Attached Sheets) OUTSTANDING QUA'LIFICATION Equipment Description S I- QUA IFI- SPECIFICATION METil0D ITEMS PARAMETER  ; QUALIFICATION

-SYSTEM: Safety Injection PERATING TIME 14 Hours 1 Day (Tabl Q6.4-1) (Vol. I p. D-1) T t " ""

j 16b PLANT ID NO.: 1-SI878A&C (p. 10) l 1A 1B Simultaneous None i-TEMPERATURE Figure 1A Figure IB (Figure Q6.4-5) (Vol. I, p, D-10 Test I

COMPONENT: Valve Motor Operators PRESSURE lA 1B Simultaneous None MANUFACTURER: Figure 2A Figure 2B (Figure 14.3.4-8) (Vol. I,. Test Limitorque/ Peerless p. D-10)

MODEL NO.: SMB / 1B Simultaneous None I Note 2 p. D-1) Test Frame PH56F/ Class B Ins. 100% 100% (Vol.

FUNCTION:

, Remotely ,

(p. 15)

Controlled Valve Operation 1. Simul taneou,

None -

Al H3B03-Na0H H3B0-Na0il 3 1B ACCURACY: Solution Solution (p. 6. -15) (V 1. I, p. D-8

2. g neering N/A R pH: 7.9- pH: 7.85 & 10) Analysis 10.0 2.30 x 2 X 103 1A . IB Seouential None 5E CE: e o" V ssel Safety RADIATION 107 Rads Ra (FigureQ6.4-4) (Vol. I, p. 5-4) Test 7n ectjg e Va1v 7% 16B (p. 2) 1B Sequential None LOCATION: Containment AGING 40 Years 40 Years (V 1. I, pp. 5-1 Test.

26'. Elevation (PP' 1-ll&l3) & 5-2)

FLOOD LEVEL ELEV.14' 10" SUBMERGENCE Note 1 ---- ---- ---- ---- None AB0VE FLOOD LEVEL YES See attached sheet for notes.

t

FACILITY: Point Beach Nucitar Plant I.-21 UNIT: 2 Rev. 5 -

00Ci'.ET: 50-301 SYSTEM COMPONENT EVALUATION WORK SHEET

~~ UDCUKtifiAUNITUUiDicB--

ENVIRONMENT (See Attached Sheets) ~

QUALIFICATION OUTSTANDING Equipment Descriptien PARAMETER S I- 00^ -

SPECIFICATION QUALIFICATI0N' METHOD ITEMS

^ " '"

SYSTEM: Safety Injection 14 Hours 1 Day (Table Q6.4'1) (Vol. p. D-1) "."* '

16B PLANT ID NO.: 2-SI878A&C b r

1A 1B Simultaneous None TEMPERATURE Figure 1A Figure 1B (Figure Q6.4-5) (Vol. I, p. D-10 Test .

COMPONENT: Valve Motor Operators 1A 1B Simultaneous None MANUTACTORER:

PRESSURE Figure 2A Figure 2B (Figure 14.3.4-f ) (Vol. I, Test i limitorque/ Reliance p.'D-10) ~

MODEL NO.: 1B Simultaneous None SMB/ Frame Tl Note 2 p. D-1) Test P56/ Class B Ins. Dl 100% 100% (Vol.

FUNCTION: Remotely . (p'. 15)

Controlled Valve Operation.

H3B03-Na0H H3B03-Na0H 1B Sequential None-CHEMICAL lA ACCURACh: N/A SPRAY Solution Solution (Vol. I, pp. D-8 Test .

pH: 7.9 , pH: 7.85 (p. 6.4'-15) & D-10) 10.0 _

2 X 108 1B Sequential None SERVICE: Reactor Vessel Safety RADIATION 2 30 x Rads IA 107 Rads' 2 X 107 (Figure Q6.4-4) (Vol I, p. 5-4) Test Injection Line Valves Rads 160 (p.2) 1B Sequential None LO,'ATION: ' Containment 1A .

26' Elevation AGING 40 Years 40 Years (po. 4.1-11&l3) (Vol . I . pp. 5-1 Test

& 5-2)16B (p. 15 i FLOOD LEVEL ELEV.14' 10" SUBMERGENCE Note 1 ---- ---- ---- ---- None AB0VE FLOOD LEVEL YES 1

See attached sheet for notes.

FACILITY: Point Beach Nuclear Plant UNIT: 1 XII -I DOCKET: 50-266 Rev. 5 -

SYSTEM COM_PONENT EVA_LUATION WORK SHEET _

DOCUMENTATION REFERENCES ENVIRONMENT (See Attached Sheets) QUAllFICATION OUTSTANDING Equipment Uescription -

N" PARAMETER SPECIFICATION QUAllFICATION METHOD ITEMS C i OPERATING TIME SYSTEM: Electrical 1A I 1 Year -----

(Table Q6.4-1) ----- -----

Note V ),

PLANT ID NO.: 1-B32 f'

'I TEMPERATURE i COMPONENT: Motor Control Note 1 ----- ----- ----- -----

None ;l'i Center i-MANUFACTURER: Westinghouse i

Note 1 ----- ----- ----- -----

None MODEL NO.: Type W RELATIVE

.. HUMIDITY Note 1 ----- ----- ----- -----

None .

FUNCTION: 480 Volt Electrical Power Dist. i Cl; M CAL ACCURACY: N/A P

SERVICE: Safeguards RADIATION 1.02 X105 Electrical Loads Rads 2A Note V LOCATION: Auxilairy Building AGING 40 Years lA 8' Elevation (pp. 4.1-11&l3) Note V Outside Charging Pump Cubicles FLOOD LEVEL ELEY. N/A SUBMERGENCE Note 1 ----- ----- ----- -----

None ABOVE FLOOD LEVEL Yes See attached sheet for notes.

FACILITY: Point Beach Nuclear Plant UNIT: 2 XII.-2 DOCKET: 50-301 Rev. 5 -

SYSTEM COMPONENT EVALUATION WORK SHE_ET_

DOCUMENTATION REFERENCES ENVIRON _ MENT (See Attached Sheets) OHALIFICATION OUTSTANDING Equipment Description b -

O l -

PARAMETER SPECIFICATION QUALIFICATION METHOD ITEMS

, i OMRATING TIME SYSTEM: Eletrical 1A l 1 Year ('lable Q6.4-1) Note V +

PLANT ID NO.: 2-B32 I TEMPERATURE '.

Note 1 ----- ----- ----- -----

None i-COMPONENT: gg- ,

c ' 91 Center ';

MANUFACTURER: Note 1 ----- ----- ----- -----

None Westinghou.

9 MODEL NO.: Type .4  !

RELATIVE 3 llVMIDiTY Note 1 -~~-- ~~~ ----- _----

None CT . 480 Volt FUg1e ical Power Dist.

ACCURACY: N/A 3 Note 1 ----- ----- ----- -----

None SERVICE: Safeguards RADIATION 3.55 X105 2A Electrical Loads Rads Note V LOCATION: .

AGING 40 Years lA Ayxilian, Building (pp. 4.1-11&l3) Note V C Elevation -

Outside Charging Pump Cubicles ~-----

Note 1 ----- ----- -----

None FLOOD LEVEL ELEV. N/A SUBMERGENCE AB0VE FLOOD LEVEL Yes ,

1 See attached sheet for notes,

~

Page 6 of 12 Rev. 5 0UTSTANDING ITEM NOTES (Continued)

T. The acoustic monitor transducers and cable connectors on the pressurizer code safety valve discharge lines have not been environmentally qualified. j This system is not required, however, to mitigate any design-basis acci- ,

dents, therefore, continued safe operation of the plant is assured. It is our intention to obtain environmental qualification data for the presently installed system from the supplier by June 30, 1982. I U. The pressurizer heaters and bolted lug-type electrical connectors have not been environmentally qualified. The pressurizer heaters are not re, quired for mitigation of postulated design-basis accidents and are not safety-related, therefore, continued safe operation of the plant is assured. The heaters provide, however, one method of controlling Reactor Coolant System pressure while achieving and maintaining cold shutdown conditions. Our preliminary evaluation indicates that the heaters and connectors should be able to survive postulated accident environment. The cable is presently environmentally qualified. It is our intention to continue evaluation of the environmental qualification of the heaters and connectors through the vendor.

V.

• The Westinghouse Type W motor control cent;ers (MCCs) are going to be analyzed by Westinghouse to evaluate the effects of radiation and aging on the materials used to construct these MCCs. The results of this analysis will be placed in Wisconsin Electric's central equipment qualification file. In the interim, operation of PBNP can continue since the radiation dose calculated using Design Basis Accident source terms are well below the values which could cause measurable degradation.

In addition, all short-term safety functions are accomplished before the MCCs are exposed to any radiation. The long-term functions (i .e. , follow-ing the initiation of ECCS recirculation) can be accomplished by improvised operator actions (i.e., LOCA and HELB accident mitigation can be accom-plished even assuming 3 failure of the MCCs in the long term).

1 1

1

.