Text

PGE-1025, Environ Qualification Program Manual
	ML20236N782
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/30/1987
From:	Walt T; PORTLAND GENERAL ELECTRIC CO.
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	PGE-1025, TAC-64233, TDW-342-87, NUDOCS 8708120107
	Download: ML20236N782 (168)
	v • d • e

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1.0 INTRODUCTION

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[b This report describes the program established by Portland General

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Electric Company for ensuring the environmental qualification of electric equipment important to sa! ety in the Trojan Nuclear Plant. j This report documents in an integrcated manner the environmental qualification review of electric equipment important to safety performed in accordance with 10 CFR 50.49. This report also supports the Updated FSAR submittal required by 10 CFR 50.71(e). Section 3.11 of the Updated FSAR incorporates this report by reference.

Environmental qualification is an ongoing process. Changes in regulatory criteria, the need to review the qualification of new and replacement equipment and the operating experience of installed equipment, and the availability of new test data and analysis con-tribute to a continual state of flux in this area. It is antici-pated that an additional amendment (s) will be made to this document i as these variables converge over time.

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In its broadest scope, qualification of equipment includes the following aspects:

(1) Environmental qualification of electric equipment importatat to safety in accident (harsh) and normal (mild) environments.

(2) Environmental qualification of safety-related mechanical equipment.

(3) Seismic and dynamic qualification of safety-related electrical and mechanical equipment.

l This report addresses the environmental qualification of electric equipment important to safety in haesh and mild environments. Other aspects of the qualification issue noted above will be addressed in

, /O later amendments to this report or separate documents.

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tv 1-1 Amendment 1 8708120107 870730 (March 1984)

PDR ADOCK 05000344 P PDR ,

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2.0 REGULATORY CRITREIA FOR ENVIRONMENTAL QUALIFICATION l O

Q This Chapter summarizes the regulatory criteria, guidelines and standards used in the environmental qualification review of electric l equipment important to safety in the' Trojan Nuclear Plant. I l

2.1 REGULATORY CRITERIA DEVELOPMENT 1

I The NRC has established criteria for environmental qualification of safety-related electrical equipment in nuclear power plants. General Design Criterion 4 of Appendix A to 10 CFR 50 establishes the general requirement for environmental qualification of safety-related equip-ment. It states in part that " structures, systems, and components important to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated !

with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents". The Trojan Nuclear Plant was I designed to meet General Design Criterion 4; however, the NRC only recently developed specific guidelines to implement this requirement for operating reactors.

l IEEE 323-1974, "IEEE Standards for Qualifying Class 1E Equipment l for Nuclear power Generating Stations", is the current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trial-use standard.

IEEE 323-1971. After substantial revision, the current version was issued in 1974. Both versions of this standard set forth generic requirements for equipment qualification, but the 1974 standard includes specific requirements for aging, margins, and maintaining documentation records that were not included in the 1971 trial-use standard. The 1974 standard was endorsed by the NRC in Regulatory Guide 1.89 for newer plants, ie, existing construction permit appli-cants. However, no Regulatory Guide was ever issued adopting the 1971 IEEE 323 standard.

L 2-1 Amendment I (March 1984)

Design of the Trojan Nuclear plant was initiated well before the issuance of IEEE 323-1971. Nevertheless, PGE made a commitment to the Westinghouse qualification program, Which used IEEE 323-1971 as a design standard for many important Nuclear Steam Supply System compo-l nents, especially those inside the containment Building where the most severe post-accident environmental conditions were postulated. l Applicable provisions of IEEE 323-1971 were also factored into the I

design of balance-of-plant equipment outside Containment where appro-priate.

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In 1978, the NRC issued IE Circular 78-08 requesting licensees to I examine installed safety-related electrical equipment and ensure I appropriate documentation of its qualification to function under accident conditions. The following year, IE Bulletin 79-01 was issued requiring licensees to provide written evidence of electri-cal equipment qualification. Although these documents requested qualification reviews and information, neither provided specific guidelines for conducting detailed qualification reviews.

1 Subsequently, in late 1979 the NRC staff developed definitive criteria for reviewing the environmental qualification of safety-related elec-l trical equipment. The Division of Operating Reactors " Guidelines for 1

Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines) were developed specifically for operating reactors. In addition, for reactors under licensing review, the NRC issued NUREG-0588, " Interim Staff position on Environmental Qualification of Safety-Related Electrical Equipment".

The intent of the DOR Guidelines is not to provide guidelines for implementing either version of IEEE 323 for operating reactors. The intent is rather to provide a basis for judgements required to con-l firm that operating reactors are in compliance with General Design Criterion 4. The intent of NUREG-0588 is to implement IEEE 323 for plants under licensing review. It provides a number of NRC staff positions on selected areas of the qualification issue. These positions are divided into two categories:

Amendment 1 2-2 (March 1984) l _ _ _ ____

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(1) Catescry I positions apply to equipment qualified in accordance with IEEE 323-1974.

(2) Category II positions apply t.o equipment qualified in accordance with IEEE 323-1971.

After the NRC staff cmapleted their initial review of licensees' responses to IE Bulletin 79-01, IE Bulletin 79-01B was issued in January 1980 requiring additional qualification information in a specified format for electrical equipment evaluated against the DOR Guidelines. NUREG-0588 was also referenced as a source of supple-mental information to be used with the DOR Guidelines.

On May 23, 1980 the NRC Commissioners issued a Memorandum and-Order-(CLI-80-21). In this Order, the Cotumission endorsed the NRC staff's use of the DOR Guidelines to review operating plants and NUREG-0588 to review plants under licensing review as well as those pieces of equipment in operating plants which do not meet the DOR Guidelines.

( The Commission directed the staff to complete its review of environ- l mental qualification, including publication of Safety Evaluation Reports (SER), by February 1,1981, and ordered that all safety-related electrical equipment in all operating plants be qualified to the DOR Guidelines or NUREG-0588 by no later than June 30, 1982.

In order to implement the provisions of the Commission's Memorandum 1

and Order, the NRC issued two separate Orders to licensees. An Order i in August 1980, subsequently amended in September, required submittal of a complete response to IE Bulletin 79-01B by November 1, 1980.

An Order in October 1980 amended the operating license of Trojan and other cperating plants to require: (a) qualification of all safety-related equipment by June 30, 1982 in accordance with the DOR Guide-lines or NUREG-0588, and (b) establishment of a central documentation file for environmental qualification by November 1, 1980. These i

requirements are documented in Trojan Technical specification 6.13.

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2-3 Amendment 1 (March 1984) 1

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on Junu 25, 1982 the Comrtission promulgated an interim rule as i 10 CFR 50.49. This rule suspended the June 30, 1982 deadline for completion of environmental qualification of safety-related electrical

, equipment pending publication of a final rule on qualifiention of such !

equipment.

On January 21, 1983 the Commission promulgated a final rule on ,

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envirorrental qualification of electric equipment important to ]

safety. This rule superseded the June 25, 1982 interim rule and was l intended to codify the methods and criteria that meet NRC requirements )

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in the area of environmental qualification.

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In June 1984, the NRC issued Revision 1 to Regulatory Guide 1.89.

j Revision 1 to Regulatory Guide 1.89 describes a method acceptable to ;

the NRC staff for complying with 10 CFR 50.49. The Regulatory Guide l generally endorses IEEE 323-1974, subject to certain clarifications and conditions.

2.2 REGULATORY BASIS FOR ENVIRONMENTAL QUALIFICATION REVIEW The review described in this report was originally conducted in accor-dance with the DOR Guidelines. NUREG-0588 was used as supplemental ,

guidance. This approach was consistent with the commission's 3 Memorandum and Order (CLI-80-21) as implemented by Trojan T2thnical Specification 6.13. This report has since been amended to address the requirements of 10 CFR 50.49. '

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i The DOR Guidelines, NUREG-0588, and 10 CFR 50.49 have been implemented 1 in accordance with the following:

(1) Equipment ordered prio to February 22,1983-(13M s effective date of 10 CFR 50.49) is qualifice in.

accordance with the DOR Guidelines or NUREG-GF60 g f (Category I)', s ,

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, t i Amenarmt 2 (Ju i N,35) 2-4 *

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(2) Equipment ordered after February 22, 1983, but before ,

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June 1, 1984 (the effective issue date of Revision 1 to D Regulatory Guide.l.89), will be qualified in'accordance with WUREG-0588 (Category I), unless the provisions of Item 4 below are applicable. Qualification in accordance with I NUREG-0588 (Category I) is construed to be equivalent to meeting the provisions of 10 CFR 50,49.

(3) Equipment ordered after June 1, 1984 will be qualified in

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th accordance with'10 CER 50.49, unless the provisions of Item 4 below are applicable. The provisions of Revision 1 to Regulatory Guide 1.89 will be complied with in the l

qualification review of this equipment, except as noted in gl PGE-1028. " Regulatory Guide Policy Manual".

1 (4) Replacement equipment ordered after February 22, 1983 must be qualified in accordance with the provisions of Items 2 and 3 above, unless there are sound reasons to the contrary. The following are considered to be sound reasons for.the use of replacement equipment previously qualified in accordance with the DOR Guidelines or NUREG-0588 in lieu of upgrading:

(a). The item of equipment to be replaced is a component of l 1

equipment that is routinely replaced as part of normal g equipment maintenance, es, gaskets, 0-rings, and coils; these may be replaced with identical components.

(b)' The item to be replaced is a component that is part of an item of equipment qualified as an assembly; these may be replaced with identical components.

(c) Replacement equipment qualified in accordance with the t provisions of 10 CFR 50.49 does not exist.

Amendment 2 2-5 (June'1985)

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l (d) Replacement equipment qualified in accordance with the provisions of 10 CFR 50.49 is not available to meet ,

installation and operation schedules. However, in such l

l case, the replacement equipment may be used only until upgraded equipment can be obtained and an outage of sufficient duration is available for replacement.

(e) Replacement equipment qualified in accordance with 10 CFR 50.49 would require significant plant modifications to acconnodate its use.

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i I (f) The use of replacement equipment qualified in accordance with 10 CFR 50.49 has a significant probability of creating human factor problems that would negatively affect plant safety and performance, n

$ for example:

1) Knowledge, skills, and ability of existing plant staff would require significant upgrading to operate or maintain the specific replacement equipment;

2) The use of the replacement equipment would Nate a one-of-a-kind application; or

3) Maintenance, surveillance, or calibration activities would be unnecessarily complex.

For replacement equipment qualified in accordance with Paragraphs (c) through (f) of this item, a nonconformance report shall be processed as described in Section 5.1.2.3.

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Amendment 2 2-6 (June 1985)

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3.0 EQUIPMENT REVIEWED FOR ENVIRONMENTAL QUALIFICATION This chapter discusses the scope of equipment considered in the environ-mental qualification program and the procedures t. sed to identify this equipment. Also discussed in this chapter are operating time require-ments for systems and equipment. )

j 3.1 SCOPE OF EQUIPMENT The equipment included within the scope of the environmental qualifica-tion program is that electric equipment defined as "important to safety" in 10 CFR 50.49(b). This equipment is defined as follows: I

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(1) Safety-related electric equipment. This equipment is that relied upon to remain functional during and following Design 1 1

Basis Events to ensure: (a) the integrity of the reactor coolant pressure boundary, (b) the capatllity to shut down the l ,

reactor and maintain it in a safe shutdown condition, and l f (c) the capability to prevent or mitigate the consequences of

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l accidents that could result in potential offsite exposures comparable to 10 CFR 100 guidelines. Design Basis Events are defined as conditions of normal operation, including anticipated operational occurrences, Design Basis Accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (a) through (c) of this paragraph.

(2) Non-safety-related electric squipment whose failure under postulated environmental conditions could prevent satisfactory secomplishment of safety functions specified in Item (1) above by safety-related equipmentI " .

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[a] Any electric equipment identified in this category is treated as lh l safety-related, as defined in Paragraph (1), and is designed and .]

l qualified in accordance with safety-related (ie, Class lE) standards l as applied to the Trojan Nuclear Plant. l l

i Amendment 2 1 3-1 '

(June 1985)

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1 (3) Certain post-accident monitoring equipment specified as j Category 1 and 2 in Regulatory Guide 1.97.

Section 3.2 describes the procedure used to identify equipment defined by 10 CFR 50.49(b). !

l The environmental qualification program also includes electric equip-ment not currently in the scope of 10 CFR 50.49. This equipment includes: (a) electric equipment important to safety in mild environ- j ments, and (b) certain post-accident monitoring equipment in harsh environments specified as Category 3 in Regulatory Guide 1.97. These i equipment categories are included in the environmental qualification program in order to ensure that all equipment qualification efforts are effectively integrated throughout Plant design. Sections 3.3 and l 3.4 describe the procedure used to identify equipment in these categories.

Environmental qualification of mechanical equipment is not presently included within the scope of this review. Notwithstanding, some qualifi- >

cation reviews have been conducted for mechanical equipment outside containment in areas that contain piping which could circulate radioactive fluids following an accident. Also, as part of the electrical equipment qualification effort, other non-electrical components, such as lubricants, j were reviewed for compatibility in harsh environments. It is intended to l

amend this report to address the environmental qualification of mechani-cal equipment once definitive regulatory criteria is forthcoming from the NRC.

3.2 PROCEDURE FOR IDENTIFYING EQUIPMENT IMPORTANT TO SAFETY IN HARSH ENVIRONMENTS The following describes the procedure used for identifying electric equipment required to be environmentally qualified in accordance with the requirements of 10 CFR 50.49.

'l Amendment 1 3-2 (March 1984)

3.2.1 IE BULLETIN 79-01B MASTER LIST DEVELOPMENT 1

I IE Bulletin 79-01B required, in part, that the NRC be supplied with a

" Master List" of all equipment required to function in a harsh environ-ment in order to mitigate a high energy line break (HELB) or Loss-of-Coolant Accident (LOCA). The steps followed to ensure a complete and accurate Master List were:

(1) Areas c' the Plant that could be subject to harsh accident environments were identified. For purposes of the IE Bulletin 79-01B response,.a harsh environment location was defined as an-area that would be subject to significantly higher than normal levels of temperature, pressure, chemical spray, humidity, or radiation as a result of a LOCA or HELB. By reviewing the original FSAR and PGE Topical' Report PGE-1004(1 , these areas were determined to be:

(a) All areas of Containment following a LOCA or an in-Containment HELB.

b (b) Main Steam Support Structure following an ex-Containment

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main steam or feedwater line break, except for areas immediately open to the outside.

(c) Auxiliary Building below Elevation 45 ft following a LOCA.

(d) Piping and electrical penetration areas following a LOCA.

Line breaks in the auxiliary feedwater pump rooms were excluded from consideration, as the redundant trains are separated from each other as well as from the effects of a main steam or feedwater line break.

(2) Safety functions needed to mitigate a LOCA or HELB were

( defined and the corresponding systems identified.

3-3 Amendment 1 (March 1984)

(3) The Plant circuit schedule was reviewed to identify safety-related equipment in these areas. The list generated by this review was reviewed to identify the safety function of each item. Only equipment needed to mitigate a LOCA or HELB accident and aubject to a harsh environment by the accident it is intended to mitigate was included on the Master List.

(4) Plant emergency procedures were reviewed to identify the equipment used by operators in response to a LOCA or HELB. The safety function of equipment identified by this review, and not already on the Master List after Step 3, was evaluated to determine the impact of each component's failure. Equipment whose failure could significantly hinder the operators' accident response was added to the Master List.

(5) Scheme drawings for all valves on the Master List wer e checked to identify stem-mounted limit switches and to determine the safety-related function, if any, of these switches. Any stem-mounted limit switch Whose failure could seriously impair response to an accident was included on the Master List.

(6) Plant status panel inputs subject to accident environments were reviewed. Stem-mounted limit switches whose failure would deny operators information essential to their accident response were included on the Master List.

(7) The circuit schedule was reviewed to determine the cable types used for representative Master List items.

I j (8) A Plant walkdown was conducted to verify equipment model numbers and locations and to identify the installed con-figuration of equipment.

i Amendment 1 3,4 (March 1984)

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, 3.2.2 10 CFR 50.49 MASTER LIST DEVELOPMENT (i

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Subsequent to the review performed in response to IE Bulletin 79-01B, a more comprehensive and systematic review was conducted to identify all equipment in the Trojan Nuclear Plant important to safety as required by 10 CFR 50.49. The updated Master List for this review was generated as follows:

(1) A listing of all channelized electrical equipment (sorted by i

system) was used to provide an initial identification of equipment important to safety. This list included all Class 1E and non-Class IE electrical equipment installed as channelized.

(2) System piping and instrument diagrams (P& ids) were reviewed to determine the safety function of each component listed on the channelized equipment list. Equipment required to remain functional during or following design basis LOCAs and HELB

( accidents was identified. The P& ids were also reviewed to

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identify any additional electrical equipment within the scope ,

of 10 CFR 50.49 not identified on the channelized equipment U list, including any directly mechanically connected auxiliary systems with electrical components which are necessary for the required operation of the safety-related equipment. All

non-safety-related components whose failure would not prevent functioning of safety-related components were eliminated from further consideration. Table 3-1 lists the safety-related functions and corresponding Trojan systems needed to mitigate l a LOCA or HELB event.

(3) The Updated FStJt and Plant Emergency Instructions EI-1, " Loss of Reactor Coolant", and EI-2, " Loss of Secondary Coolant",

l were reviewed to verify that all equipment required to mitigate the effects of a LOCA or HELB had been identified.

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3-5 Amendment 2 (June 1985)

21 (4) Electrical scheme drawings were reviewed for each safety-I related component to:

I (a) Verify that all control panels containing safety-related controls were included, and 1

(b) Identify any components (safety-related or non-safety-related) that share the first upstream breaker or fuse with the safety-related component under consideration.

3 Any components whose failure would prevent the' safety-related component from performing its safety function were included in the qualification program.

(5) A Component Summary Sheet (CSS) was developed for each elec- ;

trical equipment item identified as being important to safetyI *I. The CSS identifies the essential performance requirements of the component in question and the environ-mental conditions in which the component must function. The CSS also differentiates equipment by harsh and mild environ- ,

ment locations. Each electrical equipment item identified as

@ being important to safety as described in Items 1 through 4 ;

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above was included on the Master List. j I

i Table 3-2 is the Master List of electrical equipment important to h safety at Trojan within the scope of 10 CFR 50.49(b), and is generated from a sort of the CSS for equipment in harsh environments.

l Post-accident monitoring equipment specified as Category 1 and 2 in Regulatory Guide 1.97 is included in the environmental qualification (a) The CSS is one of two controlled documents developed as part of the environmental qualification program at Trojan. The other control-led document is the Equipment Qualification Summary. The develop-ment and use of these two documents are discussed in Chapter 5.

The CSS is discussed in this section as it relates to development of the 10 CFR 50.49 Master List.

O Amendment 2 3-6 (June 1985)

I program to the extent described in Topical Report PGE-1043( .

Because implementation of Regulatory Guide 1.97 is ongoing, not all (O),

G Category 1 and 2 equipment is currently environmentally qualified; how- g ever, certain Category 1 and 2 equipment items have been included on * '

the Master List and will be qualified in accordance with the schedules l

provided in Topical Report PGE-1043.

The approach used to review the environmental qualification of equip-ment important to safety in harsh environments is described in Sec-tion 5.1.

3.3 PROCEDURE FOR IDENTIFYING EQUIPMENT imp 0RTANT TO SAFETY IN MILD ENVIRONMENTS Environmental qualification of electric equipment important to safety located in a mild environment is not included within the scope of 10 CFR 50.49. According to 10 CFR 50.49, a mild environment is an environment that would at no time be significantly more severe than the

.<~x environment that would occur during normal plant operation, including

( anticipated operational occurrences. The criteria used to distinguish mild from harsh environment areas is delineated in Chapter 4.

1 The identification of equipment important to safety in mild environ-ments was conducted in conjunction with the identification of equipment in harsh environments. This process is described in the preceding section. The CSS documents all equipment important to safety in mild environments by component identification number, Plant location, func-tional and environmental requirements, and equipment sata.

The approach used to review the environmental qualification of equipment important to safety in mild environments is described in Section 5.2.

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3-7 Amendment 2 (June 1985)

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i 3.4 PROCEDURE FOR IDENTIFYING OTHER EQUIPMENT SUBJECT TO ENVIRONMENTAL QUALIFICATION REVIEW Regulatory Guide 1.97 specifies that Category 3 instrumentation "should be of high-quality comme'ecial grade and should be selected to withstand the specified service environ'ent".

m Generally, the specified service environment is the normal environment; however, for several Category 3 post-accident instr' ants at Trojan, it is appropriate to specify the service environment as the post-accident environment. It was also l

l considered appropriate to recategorize several post-accident instruments l l from Category 2 to Category 3 on the basis of previous requirements l specified by NUREG-0737.

The instrumentation in question concerns the post-accident sampling l

system (NUREG-0737. Item II.B.3), the noble gas effluent monitors (II.F.1.1), and the iodine and particulate sampling capability (II.F.1.2). NUREG-0737 specifies environmental qualification require-ments for the noble gas effluent monitors. This requirement appears in I

Table II..F.1-1 and states that "the instruments shall provide suffi-ciently accurate responses to perform the intended function in the environment to which they will be exposed during accidents". NUREG-0737 does not specify any environmental qualification requirements for the post-accident sampling system or iodine / particulate sampling capability.

Subsequent to NUREG-0737, 10 CFR 50.49 was promulgated and requires environment. qualifi% tion of "certain post-accident monitoring equip-ment" specified as Category 1 and 2 in Regulatory Guide 1.97. The noble gas efiluent monitors are listed in Regulatory Guide 1.97 as Cate-sory 2. The post-accident sampling system and iodine / particulate samp1-ing capability are listed in Regulatory Guide 1.97 as Category 3.

The various environmental qualification requirements for the aforemen-tioned equipment are summarized in Table 3-4. With respect to the noble gas effluent monitors, PGE has adopted the position that Regulatory O

Amendment 2 3-8 (June 1985)

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l Guide 1.97 Category 3 design and qualification criteria will be applied to this equipment. This position is justified on the basis that:

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(1) Category 3 criteria are equivalent to the NUREG-0737 criteria 13 under which this equipment was installed. I (2) Effluent radiation monitors are not necessary for the safe

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shutdown of the reactor, long-term core cooling, or containment of radioactive material following an accident.

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(3) More rigorous demonstration of environmental qualificat.,on as 13 would be required by imposition of Category 2 criteria is not l

justified by the importance of the safety function this equip-l ment performs.

l The approach used to review the environmental qualification of the fore-mentioned equipment is described in Section 5.3.

3.5 POST-ACCIDENT SYSTEM OPERATING TIMES

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U Post-accident operating times for safety-related systems and equipment were based on an AIF paper entitled, "A Nuclear Industry Position Paper on System Operating Times" . The AIF paper was submitted to the NRC for consideration on August 24, 1982.

Table 3-5 provides the generic operating time guidelines and functional requirements defined by the AIF position paper. Corresponding safety function and system / equipment requirements for Trojan, as identified from the Updated FSAR and emergency operating procedures for primary and secondary system pipe breaks, are also included.

Table 3-6 provides Plant-specific system operating time requirements for safety-related systems and equipment at Trojan. As noted in Table 3-6, individual components may not be required for the total l

i interval to ensure system operability for the required period.

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b 3-9 Amendment 2 (June 1985)

T a et E s ._1_ Sheet 1 of 3 1

SAFETY.RELATED SYSTEMS / FUNCTIONS NEEDED i TO MITIGATE A LOCA CR HELB ACCIDENT i

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SYMTEM DE9CRIPTION

\ TVPICAL SAFETY FUNCTTONS

%. N/A COMPCNENT SEAL, PANEL CABLE SEAL & PROTECTICN DNA TROJAN PLANT CABLE CCAX CABLE /PAM COMPCNENT SEAL, CCNTROL CABLE: INSTRUMENT CABLE: PCWER SUPPLY: PCWER

UPPLY, H & V SEAL ELEC CABLE 002 12SV DC BATTERY BREAKER, DC POWER DISTR, LCAD INDICATION: POWER CONVERTER . POWER SUPPLY 00+ t,16 kV' AC PCWER DISTR 005 +80.V LCAD CENTERS INDICATION & CONTROLS VOLTAGE REDUCTICN 006 480.V' MOTOR CONTROL CENTERS AC PCWER DISTRs VOLTAGE REDUCTION ..

007 LIGHTING PANEL POWER SUPPLY AC POWER DISTR, ALTERNATE DC: OC POWER DISTR: ESFAS AC LIGHT: LIGHTING, ' LIGHTING CONTRCL POWER DISTR 011 SERVICE kATER ALTN TRAIN START: BACKWASH CPERATION, CONTROLS ELEC CCNN, INDICATION & CONTROL ISCLATION SEAL IN PREVENT START PSSN TRAIN START, PUMP MOTOR: SC I/II iTRAINER MCTCR, SW PUMP TRIP; XFR ALTERNATE PUMP 1 016 COMPONENT CCCLING WATER AUTC START 110A,C AUTC START j 2108,03 CCW MKUP PUMP DRAIN)

CCW PUMP MOTOR: CCW/RNR HX ISLN CONDUCTOR, CONTAINMENT ISLN/PAM, CONTAINMNT CLR THRCT, ELEC CONN, FLOW CRIFICE BYPASS; MP 218 A CONTRCL MP 218 B CCNTROL, POWER ISCLATION (4 / PRESSURE CONTROL:

SC I/II ISLN IND, RELIEF SC I/II VALVEg ISCLATION, SI SEAL IN/IND:

TEMP IND/PAM 018 INSTRUMENT AND SERVICE AIR CI IND/PAM, CI SEAL IN/IND:

CONTAINMENT ISLN 023 DIESEL FUEL CIL SYSTEM CONTROLg. CONTROL ALARM; ELEC CONNg FUEL TRANSFER 02+ STANDBY DIESEL GENERATORS CONTRCL COOLING WATER:

FIELD WALVEg INDICATION:

NEUTRAL GND RESISTOR: +,16KV POWER SUPP!Y l 030 CONTROL BLOG H & V AIR CLG FILTRATICN

! AIR CCCLING:

CONTROL ROCM ISLNg CONTROL ROOM PRESS: ELEC CCNN, M2 EXCHANCE, INDICATICN & CONTRCL ROOM. HEATER 032 FUEL / REACTOR AUX BLOGS H & V CI SEAL IN/PAM: CONTAINMENT CONTAINMENT ISLN/PAM:

ISLN ELEC CONN i

INDICATION & CONTROL: I ROOM AIR CCCLING j 033 TURBINE / AUX BLDGS H & V INDICATION & CONTRCLs RCCM j AIR EXHAUST, ROOM AIR SUPPLY: 1 ROOM CCCLING SEE NOTES 03+ MISCELLANEOUS BLOGS H & V ROCM CCCLING 035 FUEL PCCL CCCLING & DEMINERALIZATION LEVEL IND/PAM SC I/II ISLN INDs NOTES SC I/II .ISCL ATION: SEE 0+5 FEEDWATER SYSTEM AFW DIESEL BATTERY, AFW DIESEL CONTRCL: AFW INLET VALVE:

AFW ISCLATION: AFW SUPPLY:

BACK UP FDW SOURCEg CCNTROL:

CONTROL PUMP SPEED, CONTRCL VALVE: DIESEL AUTC START:

ELEC CCNN: ELEC CCNN/ISLN3

[\ FDW BYPASS ISCLATION FDW ISCLATICN, FLCAT BTRY CHARGEg FLCW IND/PAM: IND CPEN VALVE:

Amendment 3 (December 1985)

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11/20/as T a aL E s.1 Sheet 2 of 3.

SAFETY.RELATED SYSTEMS / FUNCTIONS NEEDED TO MITICATE A LOCA CR HELB ACCIDENT Sv97EM DESCRIPTION TYPICAL SAFETV FUNCTTCNS i 0+5 FEEDWATER SYSTEM IND SHUT VALVE: INDICATION l & CCNTROL LOW CCCLING WATER:

l LOW PRESS TRIPi CPERATOR SPEED l INCRg PUMP MOTOR: REACTOR TRIP, SC I/II ISCL ATION:

SIGNAL ISCL ATION: SPEED INCREASER; STEAM SUPPLY, THRCTTLE VALVE:

TRIP DIESEL PUMP 3 TRIP TURBINE PUMP 0+9 RESIDUAL HEAT REMOVAL CONDUCTOR: CONTAINMENT ISLN/PAMj j ELEC CONN: ELEC CONN /PAM i

FLOW CONTROL, FLOW IND/PAMj l

FLOW INDICATION: INTERL CCK MINIMUM FLOW ISLN MC 881+

INTERLCCK PAM PUMP MOTCRs RADN WASTE ISCLATION: RMR ISLN/PAMg RHR ISCLATION:

RHR SUMP ISCL ATION: SI SEAL ,

IN/IND: TEMP INDICATION: }

TRAIN ISCLATION j 050 CHEMICAL & VOLUME CONTROL CCP LOW PRESS START: CCP CCP PUMP MOTCR, l l PUMP LUB CIL:

CI SID/PAM: CCNDUCTOR: CONTAINMENT ISLN CONTAINMENT ISLN/PAM:

CONTRCL, IND & CCNTRCL/PAM INDICATION & CONTRCL ISCLATION:

MIN FLOW LINE ISLN, PAM RADN WASTE ISCLATION: VCT ISCLATION 051 HEAT TRACING ELEC CONN /PAM: FREEZE PROTECT /PAM:

IND & CCNTRCL/PAM 052 SAFETY INJECTION BIT ISCLATION: CI IND/PAh, CONDUCTOR; )

CI SEAL IN/PAM:

CONTAINMENT ISLN: CONTAINMENT ISLN/PAM: CONTROL, CVCS ISCLATION ELEC CONN ELEC CONN /P4Mi FLOW INDICATION: ISCLATION: 4 PAM; PUMP MOTOR: RADN WASTE )

ISCLATION: REACTOR PROTECTION:

RHR ISLN/INTERLCCK3 RWST ISLN/INTERL.

SC I/II ISLN IND: SI ALIGNMENT:

SI SEAL IN/IND 053 ENGINEERED SAFECUARDS ACTUATING CONTROL, ESFAS: INDICATION

& CONTRCL3 PAM l 05+ SEISMIC MONITORING SEE NOTES 056 MEACTCR CONTROL & PROTECTION CONTROL, CONTROL & CPERATION:

INDICATION & CONTROL 057 120.V PREFERRED INSTRUMENT AC AC POWER DISTR POWER INVERTER 059 PRIMARY CONTAINMENT CONTAINMENT ISLN.

060 CONTAINMENT H & V CI IND/PAM CONDUCTOR: CONTAINMENT AIR CLC CONTAINMENT M2 MXC:

CONTAINMENT ISLN, CONTAINMENT ISLN/PAM: CONTRCL, ELEC CONN 3 HE CONTRCL3 IND & CONT /PAM:

IND & CONTROL /PAM: INDICATICN, INDICATICN/PAM: PAM, POST ACCIDENT SAMPLEj PCWER SUPPLY 061 CONTAINMENT SPRAY CSS PUMP MOTOR DISCHARGE IND/P AMr ELEC CCNN: INDICATION

& CONTRCL; LEVEL IND & ALARMS NACH TANK ISOLATICN RADN WASTE ISCLATICNg SPRAY ISCLATION:

SUMP ISCLATION 063 STEAM CENERATORS CONTAINMENT ISLN, PAM: PAM/RSS:

STEAM LINE ISLN IND STEAM LINE ISCLATION 06+ RCS INCLUDING PRESSURIZER CI IND/P AM; CI PASS /PAM:

CONTAINMENT ISLN, CONTAINMENT ISLN/PAM; CCNTROL ELEC CONN /PAM; H2 VENT: INDIC ATION: INDICATION

& CONTROL PAM PCRV BLK 067 PRIMARY MAKEUP WATER \

CI SEAL IN/IND/PAM CONTAINMENT ISLN: LVL IND & CCNT/PAM Amendment 3 December 1985) a 8 -

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CONN /PAM: H2 VENT, INDICATION l

& CONTROL, PAM, RSS, RSS SI: SI, STEAM LINE ISLN 083 MAIN STEAM CONTAINMENT ISLN, CONTRCL j ELEC CONN: PT AM: SI/PAM: i SI, STM LN ISLN/PAM: STEAM I LINE SEAL IN ISLN IND: STM LN ISLN i 099 MISCELLANEOUS SYSTEMS INDICATION & CONTROL: SICNAL ISCLATION VCLTAGE RDCN, CH I As VCLTACE RDCN, CH B l

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E B 4 4 4 4 H M 4 4 4 4 4 4 4 E St 1 T E 4 4 4 4 4 H

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5 2

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2 9

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4.0 SPECIFIED ENVIRONMENTAL CONDITIONS AND EVALUATION CRITERIA

%./

This chapter describes the environmental conditions for which equipment must be qualified in order to perform its safety function in normal and accident environments. The criteria used for evaluating the qualifi-cation of equipment to specified environmental service conditions are also described. Unless otherwise noted, these criteria are applicable !

for electrical equipment reviews conducted to both DOR Guidelines and NUREG-0588 requirements.

I Mild and harsh environment locations in the Plant are differentiated on j the following basis: !

(1) All equipment located in an area subject to the effects of high !

energy pipe breaks, or where the total integrated radiation dose (1.1 x accident Y and B, plus 40-yr background) is

I 3

>10 rads, is considered to be in a harsh environment.

(2) All remaining equipment is considered to be in a mild b environment.

t Table 4-1 lists the mild and harsh environment conditions for each plant location.

4.1 TEMPERATURE AND PRESSURE 1

The following discusses the temperature and pressure profiles specified l for pipe breaks in high energy fluid systems (temperatures and/or pres-sures >200*F and 275 psig, respectively) inside and outside containment.

Normal environment temperature ranges are also summarized.

f e

A ]

4-1 Amendment 2 (June 1985)

_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l

--- . - ~ _ -

)

4.1.1 PIPE BREAKS INSIDE CONTAINMENT 4.1.1.1 Loss-of-Coolant Accident The basis for defining the environmental conditions for qualifying equipment inside Containment is provided by the containment transient i analysis performed specifically for Trojan. This analysis is described I i

in Section 6.2 of the Updated FSAR. Figures 6.2-91 and 6.2-109 of the Updated FSAR summarize the pressure and temperature conditions vs time for the worst-case Loss-of-Coolent Accident (LOCA) event. These profiles are included in this report as Figures 4-1 and 4-2.

The profiles identified in Figures 4-1 and 4-2 were used as a basis for deriving " qualification requirement profiles" for Containment pressure and temperature. The qualification requirement profiles are used to evaluate the qualification of all new (or replacement) equipment.

Figutas 4-3 and 4-4 depict the qualification requirements for l Containment pressure and temperature.

The approach above is consistent with Section 4.1 of the Division of Operating Reactors (DOR) Guidelines, which recommend that Containment temperature and pressure conditions as a function of time be based on FSAR analyses. The Trojan Updated FSAR analysis, which uses the Bechtel COPATTA computer program, a derivative of the CONTEMPT code, is also consistent with Section 1.1 of NUREG-0588 and has been previously accepted by the NRC .

4.1.1.2 Main Steam Line Break With respect to the environmental conditions resulting from a main steam line break (MSLB) inside Containment Section 4.2.1 of the DOR Guidelines states that:

l " Equipment qualified for a LOCA environment is considered I

qualif!ad for En MSLB accident environment in plants with ,

automatic spray systems not subject to disabling single Amendment 2 4-2 (June 1985)

1 i k

l 1 j component failures. This position is based on the 'Best Estimate' calculation of a typical plant peak temperature and I pressure and a thermal analysis of typical components inside Containment * . The final acceptability of this approach, ie, use of the 'Best Estimate *, is pending the completion of Task Action Plan A-21, " Main Steam Line Break Inside I Containment".

l Because Trojan is equippsd with an automatic Containment Spray System not subject to a disabling single component failure, the LOCA environ-mental conditions discussed above are considered to effectively bound I the MSLB conditionn. The Trojan containment Spray System is described !

in detail in Section 6.5.2 of the Updated FSAR. The analysis provided in Table 6.5-2 of the Updated FSAR provides sufficient verification that the Containment Spray System is not subject to disabling single component failures.

p Final acceptability of the approach stated in Section 4.2.1 of the DOR

( Guidelines (ie, use of the "Best Estimate") depends upon the resolution of Task Action Plan A-21. However, an analysis conducted for PGE evalu-ated the thermal capability of typical equipment inside containment exposed to an MSLB environment . This analysis shows that a superheat condition (with corresponding high Containment atmosphere temperature) should have no significant effect on electrical equipment temperatures because the expected equipment surface temperature would follow the Containment saturation temperature, which is substantially lower than the peak vapor temperature during the superheat phase of the accident. The reason for this is that energy transfer from the Contain-ment atmosphere to heat sinks is significant only when the sink surface is cooler than the saturation temperature so that condensation can occur.

If the equipment sutface temperature were to become higher.than saturation, then the low energy tiensfer mechanism of convection would govern heat

[a] See NUREG-0458(4) for a more detailed discussion of the best estimate calculation.

Amendment 2 4-3 (June 1985)

.m _.

transfer. Since the Containment peak pressure is at a maximum following ;

the design basis LOCA, the Containment saturation temperature for an MSLB accident is no higher than would be the case for a LOCA. It is on ,

these bases that the worst-case environment for Containment equipment will result from the design basis LOCA. Therefore, the position as stated above (Section 4.2.1 of the DOR Guidelines) should remain a valid conclusion for the Trojan design.

With regard to temperature stratification effects follow'.ng a MSLB (or LOCA), any such effects will be short-lived, ie, less than several minutes, due to Containment Spray System operation.

4.1.2 PIPE BREAKS OUTSIDE CONTAINMENT The environmental effects of high-energy line breaks (HELBs) outside !

Containment have been analyzed and are described in Topical Report PGE-1004, " Trojan Nuclear Plant Analyses of Pipe System Breaks, Outside Containment" .

A main steam or feedwater line break in the Main Steam Support Structure (MSSS) area produces the most severe temperature and pressure conditions to safety-related equipment outside Containment. Temperature and pressure effects from pipe breaks outside Containment to safety-related l equipment in other Plant locations can be eliminated from consideration since, as shown in PGE-1004, safa Plant shutdown would not be affected.

For the initial review of equipment qualification in response to IE Bulletin 79-01B, the environmental parameters for temperature and pres-sure identified in PGE-1004 were used. PGE-1004 identified a temperature and pressure peak of 240'F and 10 psig for breaks in compartments con-taining the main steam and fesdwater lines. While some safety-related l equipment is located in these compartments, most other safety-related g equipment in the MSSS is contained in lower compartments which inter-

~

connect to the upper compartments via hatch openings. Consequently, a further analysis was performed using a RELAP4/ MOD 5 computer model of the ,

MSSS to better define the environmental conditions resulting from I

Amendment 2 4-4 (June 1985)

an MELB event in this area. The results from this analysis predict b maximum temperatures and pressures of 228'F and 5.5 psis, and 212*F and 0.0 psig in the pipe break and adjacent compartments, respectively.

These conditions were used as a basis for performing a detailed environ- l mental qualification review of equipment important to safety in the MSSS. This review, and a detailed justification for continued operation for equipment whose qualification cannot be demonstrated by type test-ing, was provided to the NRC'in a letter dated June 1, 1984 (Attachment 6 thereto).

Subsequent to the review above, Westinghouse notified PGE of a possible I unreviewed safety question concerning the temperature envelope require-ments on the environmental qualification of equipment outside Contain-ment for HELBs. Analyses performed by Westinghouse show that steam generator tube bundle uncovery may occur during an MELB, resulting in superheating of the stcam exiting from the steam generator. This effect 1

results in an increase in the temperature of the steam. Westinghouse I l provided guidelines for use by Westinghouse Owners Group (WOG) member g a utilities in performing an interim evaluation of the effects of super-heat steam blowdowns on equipment outside containment. Included with these guidelines were sample mass / energy releases for breaks outside Containment ranging from a large double-ended rupture to a small split l 1

rupture. These mass / energy releases are based on a four-loop plant.

Conservative assumptions were made in order to result in early tube bundle uncovery and, therefore, the earliest superheat initiation time.

Although the sample mass / energy releases are neither generic nor neces-sarily conservative for a given plant, they are representative of the superheat steam phenomena and provide a valid estimate of the effects on compartment temperature analyses.

The sample blowdown data received from Westinghouse was input to the RELAP4/ MOD 5 model of the MSSS. The results show a considerable increase in the temperature response of most compartments in the MSSS from those temperatures predicted by the earlier analyses referred to above; how-g ever,'the predicted temperature and pressure responses in the non-break

(\

compartments of the MSSS are bounded by the Containment requirement profiles (Figures 4-3 and 4-4).

4-5 Amendment 2 (June 1985)

p .v _.

q l

Interim pressure and temperature requirement profiles for qualifying replacement equipment in the MSSS are provided in Figures 4-5 and 4-6.

Confirmation of the adequacy of these interim profiles is pending

7 receipt and analysis of Plant-specific blowdown duta from Westinghouse.

Plant-specific blowdown data is currently under development via a WOG l subgroup program. This program is scheduled for completion by September 15, 1981. Establishment of final temperature and pressure profiles is scheduled by December 15, 1985.

4.1.3 NORMAL ENVIRONMENT TEMPERATURES Table 4-1 summarizes the normal (mild) temperature and humidity ranges specified for various Plant areas. This data was used for aging analysis f of harsh environment equipment and is also included for future qualifica-tion review of safety-related electrical equipment in mild environments. j 4.1.4 EVALUATION CRITERIA 4.1.4.1 Temperature The equipment is considered qualified for temperature effects if one of the following conditions is met:

(1) The equipment type test temperatui ,cofile encompasses the specified temperature profile.

(2) The equivalent thermal degradation of the test profile referenced to 120*F is equal to or exceeds that of the specified profile [see Section 4.5.2 for discussion of the application of Arrhenius methodology to the determination of thermal degradation represented by Design Basis Event (DBE) test profile) and: (a) the peak temperature of the test profile meets or exceeds that of the specified profile, or (b) there are no anticipated changes in the electrical characteristics of the equipment for tempera-tures up to the maximum value of the specified profile. )

Amendment 2 4-6 (June 1985)

((I' (3) An analysis shows that the equipment is not susceptible to l thermal degradation, and structural deformation is not anticipated from the peak specified temperatures (applicable to DOR Guidelines only). C (4) Equipment currently installed in locations that do not i experience significant temperature changes as a result of the accident are considered to be qualified for their temperature by successful operating experience. A.1.4.2 Pressure The equipment is considered qualified for pressure effects if one of the following conditions is met: (1) The equipment type test pressure profile encompasses the specified pressure profile. 1 s ((s-'/ ., (2) The test profile meets or exceeds the maximum specified pressure at a time when temperature and humidity are above this specified maximum and spray is present. (3) The equipment function is not significantly affected by differential pressures and pressure is not a motive force for failure of the equipment being evaluated,'eg, j the component is hermetically sealed and structural failure resulting from differential pressures is not anticipated (applicable to DOR Guidelines only). Ifh 4.2 HUMIDITY A.2.1 PIPE BREAKS INSIDE AND OUTSIDE CONTAINMENT Pipe break events inside and outside Containment are assumed to produce 100-percent humidity conditions in the immediate area of the break.

                                                                                                                                                          ' Amendment 2 4-7        (June 1985)

__ _ __-_.____ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ = _ _ _ _ _ _ _ . . _ .

In some cases, electrical equipment qualification testing of existing

Plant equipment was not conducted in a manner that assured attainment of 100-percent relative humidity nor in a manner that allowed calcula-tion of actual test humidity conditions. However, the testing did expose >

components to saturated steam conditions for an extended period of time. This is judged to provide reasonable assurance i.! equipment operability under high-humidity condition because: 1 l (1) Continued injection of steam into a closed test chamber eventually resulted in an extremely high humidity condition. (2) Injection of steam into a closed test chamber is analo-gous to high-energy water / steam pipe ruptures inside Containment. 1 (3) Sufficient moisture exists in the high-temperature saturated steam conditions to identify significant component failure modes that might result from condensation or water absorption. For qualification testing of new equipment, consideration will be given to one of the following for assuring a moisture-saturat=J environment: (1) The methods described in Appendix C of IEEE 323-1974. (2) 100-percent relative humidity, or saturated conditions, at the maximum temperature and pressure achieved. 4.2.2 NORMAL ENVIRONMENT HUMIDITIES Table A-1 summarizes the normal humidity ranges specified for various Plant areas. G' ! Amendment 2 4-8 (June 1985) L-________________________________-_--_______ __ _ _ _ . _ _ _ _ . _ _ _ - _ _ _ __-

. G) l 4.2.3 EVAiJATION CRITERIA The equipment is considered qualified for humidity effects if one of the ! following conditions is met: I (1) The equipment type test humidity qualification exceeds l I the specified humidity lovel. l (2) Analysis or operating experience shows equipment per-l formance is not subject to degradation in a high-humidity environment, eg, the component is hermetically sealed or

l has successfully operated under high-humidity conditions l 1 during normal Plant operation (applicable to DOR ^ j l Guidelines only). I l I (3) The equipment type test exposed the equipment to saturated steam conditions for a prolonged period.

(/N.)1 n l (4) An analysis shows that the equipment is similar to 13 { equipment that meets one of the preceding criteria. l l (S) Existing equipment located in humidity environments that 1 do not significantly change as a result of an accident are considered to be qualified for the humidity condi-tions as a result of successful operating experience. I 4.3 CHEMICAL EFFECTS 4.3.1 PIPE SREAKS INSIDE CONTAINMEE The worst-case chemical spray conditions resulting from a LOCA event are boric acid, pH 4.2, and sodium hydroxide, pH 10.5. These bounds envelop conditions resulting from a Reactor Coolant System (RCS) pipe rupture, followed by containment Spray System operation. ( Amendment 1

                                                                                     '~9 (June 1985)

An acidic condition would result only from a ruptured RCS pipe during -' and immediately after boron injection by the centrifugal charging pumps. In this case, the corrosive effect of such a weak acid existing for such a short duration has a negligible impact on equipment. Therefore, pH effects <7.0 are not considered in the qualification program. Equipment inside Containment will be subject to Contain:nent spray with pH between 8 to 10.5 as' discussed in Section 6.5.2.3 of the Updated FSAR. During the injection phase of an accident, the pH of the spray solution is maintained between 9.0 and 10.5. Durins the long-term i recirculation phase, the recirculation sump-spray solution is main-tained between 8.0 and 9.5. Equipment inside Containment must either be tested in a basic spray environment or be analyzed for the cor-rosive effects of the spray solution. The concentration of caustics or other chemicals used for qualification should be equivalent to or more ! severe than those used in the Plant Containment Spray System. A Containment spray solution with a pH of 10.5 is a worst-case condi-tion for a short time duration (ie, <30 min), based on bounding values l of parameters (see Updated FSAR Figure 6.5-7). Actual pH values would be in the 9.0 to 10.0 vange. During the long-term recirculation phase, which is more representative of the environmental conditions equipment should be qualified to, pH is maintained between 8.0 to 9.5. A maximum pH value of 9.5 is considered a more appropriate specification for environmental qualification purposes than a pH value of 10.5 because: (a) specifying a maximum pH value of 9.5 assures that tnaterials sen-sitive to caustic or chemical spray environments will be eliminated from use or qualified for use, and (b) it is unlikely that materials qualified for long-term use (ie, days) at a pH value of 9.5 would experience significant degradation at an elevated pH of 10.5 for

           <30 min.

S' Amendment 2 (June 1985) 4-10

t 1 l t i

 //                                                                                                                                       \

([ m\ In some cases, spray testing was not done at a pH level as severe as ! the most basic Containment spray possible. This equipment is consi-dered to be qualified for its application if materials known to be reactive with the chemical spray were excluded from the equipment, and l if no failures resulted from the exposure of the equipment to the basic , pH (usually pH 8 or above) conditions that these items were subjected l to in their respective test programs. l 4.3.2 PIPE BREAKS OUTSIDE CONTAINMENT

l Chemical spray effects to equipment outside Containment from pipe ! 1 breaks are not a consideration. : l 4.3.3 EVALUATION CRITERIA The equipment is considered qualified for chemical spray effects if one of the following conditions is met: j 7

  ' /mi i                                                                                                                                      }

, U (1) The equipment type test pH qualification exceeds the 9 specified pH level. (2) A material analysis shows the equipment is not sus-ceptible to chemical degradation (applicable to DOR gj Guidelines only), *l l (3) An analysis shows that the equipment is similar to Ib equipment that meets one of the preceding criteria. j 4.4 RADIATION l 1 4.4.1 RADIATION DOSES i 4.4.1.1 Accident and Normal Radiation Dose Estimates l

    -   Post-accident Y and 8 total integrated doses (TIDs) were calculated                                                               j j  for Trojan using the methodology outlined in Appendix D of NUREG-0588                                                             j 4-11                                       Amendment 2                                       I (June 1985) l j

l i and NUREG-0578. Appendix 8.A describes these calculations. Included in ! Appendix 8.A are two component-specific sample calculations (one for inside and one for outside Containment), and a brief description of each of the methodologies used, their application and associated conservatism. Also included in Appendix 8.A is a comparison of the methodology of this report with the methodology which was used prior to its issuance. i The TIDs were based on two bounding conditions and are used for TMI-related modifications as well as IE Bulletin 79-OlB electrical equipment qualification evaluations. The first bounding condition assumed a LOCA with 100-percent core noble gases, 50-percent core halogens and 1 percent other core fission products released into the Containment. The second condition assumed the RCS remained intact, with the same source term as above contained within the RCS. The TIDs were determined for several dose point locations, both inside ard outside of the Containment. In addition, doses were calculated for equipment operation times ranging from i he to 1 ye, in orcar to determine qualification levels for the maximum time equipment is required to function following an accident. Maximum Y TID values for general areas at Trojan are summarized in ! Table 8A-1 and Figures 8A-1 through 8A-10. These values are conservative estimates based on contact dose rates from the largest sources in each area. Beta TIDs for materials in contact with reactor coolant and materials exposed to the containment atmosphere are also provided in Table BA-1. Figures 8A-ll and 8A-12 provide volume correction factors used to determine B integrated doses for finite volumes in the Containment i 1 atmosphere. l l l 1 The 40-yr backbround radiation TIDs are presented in Table 8A-1 for areas inside and outside Containment. These values are conservative estimates based on the highest expected sources in each area. f 9g l Amendment 2 4.12 (June 1985) J l - i

l l 1 The overall TID to a component was determined by summing the normal background dose for 40 yr and the y and 8 accident doses at that lccation for the assumed time the component is required to operate post accident. Appendix 8.A outlines the method for specific component calculations and contains example dose calculations for typical locations inside and outside the Containment.

                                                                                        'i 4.4.1.2   Radiation Damage Threshold Levels Because different materials exhibit a wide range of tolerance to radiation damage, radiation threshold levels have been established to exempt certain equipment from demonstration of its radiation qualification by testing or j         analysis. The bases for establishing threshold levels are discussed below; l

selected threshold values are given in Section 4.4.2. References 6 theough 8 provide surveys of radiation threshold effects on ' l l (, organic materials and electronic devicer. Because inorganic and metallic materials are much less susceptible to radiation damage they are not ; separately considered as part of the qualification review. ! l 1 Information presented in Reference 6 concerning organic materials used in Plant equipment suggests an exclusion from test or further analysis should be allowed for nonelectronic equipment subjected to 10 rads or less. Nonelectronic equipment which contains no Teflon and is subjected to 4 5 i

         <10 rads should likewise be excluded. At these levels, there is no 1

i significant degradation of mechanical or permanent electrical properties of j i the listed materials. Also at this level, no indications were found in the j literature search of significant synergistic effects of radiation combined j i with other environmental stresses or sensitization to subsequently imposed i stresses. In general, equipment failures occur at some higher level involving considerably more than threshold degradation of component materials. l i Amendment 2 i (June 1985) ! l

Reference 7 essentially confirms the results of Reference 6. Although four materials in addition to fluorocarbons (Teflon) were identified as having a change threshold between 10 -10' rads, the 25-percent damage dose level l for these materials ranged above 10 rads. For Teflon, the 25-percent h damage level is quoted in the range of 3 to 5 x 10 rada. For use as an evaluation criterion, the 25-percent damage level is considered more j indicative of radiation damage than the lowest reported change threshold l l because the changes are real at this level and are not an artifact of the i j experiment or due to sample variation. L Chapter 5 of Reference 8 summarizes the radiation tolerances of various electronic components. Guidelines are also presented for component 3 selection. Radiation damage thresholds are quoted as low as 10 rads for some electronic devices such as metal-oxide semiconductors. These b guidelines were used in establishing evaluation criteria for radiation qualification of such devices. {t 4.4.1.3 Radiation Dose Rate Effects O

To facilitate the use of a reasonable test time, an accelerated exposure rate may be necessary. Thus, to allow margin for these effects, a greater total dose than the service lifetime dose should be applied.

Current materials aging methodology used for nuclear qualification, as given in IEEE 383-1974 (applicable to electric cables, field splices, and connections), specifies an upper limit of 1 x 10 R/hr on the dose rate for radiation aging. 4.4.2 EVALUATION CRITERIA The equipment is considered qualified for radiation effects if one of the I following conditions is met: (1) The specified TID, accident (Y + 8) plus 40-yr background, is

g equal to or less than: O' Amendment 2 4,14 (June 1985)

i j (a) 10 rads for unhardened electronic devices such as SCRs, . power transistors, CMOS, NMOS and MOSFET. " (b) 10 rads for unhardened electronic devices such as ! n i bipolar transistors, operational amplifiers, resistors, 5 capacitors, diodes, and LEDs, j (c) 4 x 10 rads for nonelectronic equipment containing IQ Teflon.

                    -(d)   10 rads for nonelectronic equipment containing no l{

Teflon. l (2) The equipment type test TID qualification exceeds the specified TID. The testing must be conducted in an air environment with radiation aging dose rates $1 x 10 R/hr. Radiation aging Ih testing done at higher dose rates must be justified for ( radiation dose rate effects. (3) A material and design analysis shows that the equipment is not susceptible to radiation degradation at the speci-fled TID levels (applicable to DOR Guidelines only). Ih (4) Where radiation is the only environmental concern and the specified TID is $10 rads, a material analysis shows that the equipment is not susceptible to radiation degradation at b the specified TID levels (equipment qualified under this criterion is designated " harsh-exempt"). (5) An analysis shows that the equipment is similar to equipment Ih that meets one of the preceding criteria. Shielded components need be qualified only to the gamma radiation envi-ronment, provided it can be demonstrated that the sensitive portions of the component or equipment are not exposed to significant beta radiation (\ dose rates. If, after considering the appropriate shielding factors, the 4-15 Amendment 2 (June 1985) l (

total beta radiation dose contribution to the equipment or component is - calculated to be less than 10 percent of the total gamna radiation dose to which the equipment or component has been qualified, then the equip-ment or component is considered qualified for the beta and gamma radia- . 1 tion environment. 4.5 AGING i Aging effects were explicitly considered for all safety-rslated equipment I at the Trojan Nuclear Plant potentially exposed to harsh environments. The following approach is used for evaluating aging effects on existing equipment and discusses the degree of conformance to the requirements of Section 7 of the DOR Guidelines. l l1 4.5.1 BASIS AND APPROACH The DOR Guidelines and NUREG-0588 require consideration of aging effects on safety-related electrical equipment, but by different methods. l Section 7 of the DOR Guidelines, as supplemented by Generic Letter No. 82-09, considers the following to provide an acceptable basis for addressing equipment aging if a specific qualified life cannot be demonstrated as required by IEEE 323-1974 (Category 1 of NUREG-0588): (1) Conduct an equipment material evaluation to ensure that no known materials susceptible to degradation because of l l aging have been used, i I (2) Establish an ongoing program to review Plant surveillance and maintenance rect,rds in order to l identify equipment degradation walch may be age related, and/or (3) Propose a maintenance program and replacement schedule 6 for equipment identified in Item 1 or equipment that is qualified for less than the life of the Plant. Amendment 2 , 4-16 (June 1985)

l l

t, In evaluating the qualification of most safety-related electrical equip-ment at Trojan, the foregoing DOR Guideline method is employed only when a specified qualified life cannot be demonstrated for equipment by an accelerated aging test. Because NUREG-0588 considers the Arrhenius methodology an acceptable method of addressing accelerated aging ) 1 l (although other aging methods, as detailed in Reference 9, that can be ! supported by type tests may be used on a case-by-case basis if properly l justified), this method was used to establish a qualified life for equipment with ava'ilable pre-aging test data. The Arrhenius method was ! also used to evaluate the equivalent thermal degradation imposed by the l i test program in order to determine if materials susceptible to thermal ] 1 degradation had been used. The extrapolation of LOCA test data using an ' Arrhenius tochnique to extend a test duration or establish a qualified life is justifiable since the DOR Guidelines state "A shorter test duration may be acceptable if specific analysis are provided to demonstrate that materials involved will not experience significant accelerated thermal aging during the period not tested." And as indicated above, NUREG-0588 states "The Arrhenius methodology is [ y,/ considered an acceptable method of addressing accelerated aging." pCE considers the approach taken to be in accordance with NRC-provided guidance. In cases where insufficient accelerated aging was applied as part of an equipment test program, material analysis was conducted in accordance with Item (1) above. As an example, aging effects for Limitorque motor operators were evaluated by a combined approach using component matorial analysis and the Arrhenius methodology. Each component was evaluated as to its potential for thermal degradation, and if found to be susceptible, its qualified life was determined by Arrhenius calculations. InFulation, switches, seals, terminal blocks and lubricants were included in the material evaluation. The terminal blocks used in the motor operators are an example of qualification by material analysis. These blocks are molded phenolic, which has no potential for significant thermal degrada-tion over a 40-yr period, t ) J

Amendment 2 (June 1985)

                                                                                    'l I

l

l On the other hand, an RH-insulated motor is an example of an item l l susceptible to thermal degradation. Based on analysis of test data using the Arrhenius methodology, Limitorque gives the life of an RH-insulated motor as 9 x 10 he at 120*F. This is much greater than ! the required 40-yr life (3.5 x 10 hr). I 1 4.5.2 APPLICATION OF ARRHENIUS METHODOLOGY ' Aging theories are based on material degradation due to stress factors. The Arrhenius model simulates accelerated thermal aging by increasing j the reaction rate of certain stress factors. Enduring an environment of high stress reaction rates for short intervals of time causes degrada- '] tion equivalent to enduring an environment of lower stresses for longer intervals of time. The Arrhenius model also relates rate of reaction to ! temperature by the function: l r = A exp(-$/kT) (4-1) where O l e = rate at which the degradation reaction proceeds A = constant for the material (frequency factor)

              & = activation energy (eV) 31           k = Boltzman's constant (0.8617 x 10' eV/*K)

T = absolute temperature. *K. The following rolstionship is used to determine intervals of equivalent thermal degradation: Il \ in (t,/t,) =p$ 1 (4-2)

                                                    \s99) a Amendment 2                          4_1g (June 1985)

I where-t s= simulated or service time interval associated with T s T, = operating ce reference temperature. *K t = accelerated aging time interval associated with T, I T, = accelerated temperature. *K. Manipulating Equation 4-2 yields: equivalent " a ** ~- (4-3) I, where t y,y = time interval at operating or reference temperature ( ) for which thermal degradation is equivalent to 4 accelerated aging interval (t,) at accelerated temperature (T,). l To perform the aging evaluation of a specimen, as detailed in Section 4.5.4, four separate applications of Arrhenius methodology must be performed. Equivalent thermal degradation time (tequivalent' ""# be determined for: i (1) Pre-DBE accident aging time to determine equipment qualified life. l L (2) DBE test profile to ensure equivalent thermal degradation of test profile is equal to or exceeds the qualification requirement temperature profile (see Figure 4-4). (3) Qualifleation requirement tempmenture pf61116 (iee Figure 4-4) for comparison in (2) above. (4) Post-accident operating time. 4 19 Amendment 2 (June 1985)

                     -                                    ,    m...   -           ,

j l l l i To illustrate use of the' Arrhenius method, consider the following example

where post-accident operating time is determined. Assume the required -

post-accident operating time is 1 yr at 120*F (322*K). A conservative

representation of the example specimen DBE test profile is given by: ' ! L Inte val, t, Temperature T, 0.5 hr 250*F (394*K) 1.4 hr 120*F (322*K) 24 hr 250*F (394*K) ; 1000 hr 220*F (377*K) l Using a conservative activation energy & = 0.5, and applying the criteria of Section 4.5.2.1(3) below, post-accident operating time is ; ! determined from plateau portion of above curve (220*F): l l l _ . equivalent = 0 k en ~ _.8617 1 ev/*K _

                        = 13,859 he at 120*F This far exceeds the required post-accident operating time of 1 year (8760 hr).

Similar calculations are used to determine equipment qualified life and equivalent thermal degradation of the DBE test profile.

4.5.2.1 Selection of Activation Energy for Use In Arrhenius Methodology Limiting activation energies provided by the equipment vendor or test laboratory are used in extrapolation for pre-aging tests or post-accident operating times. When activation energies are not provided by l l equipment vendor, the selection of activation energies should be based upon a detailed review of oven-aging data for the materials involved in

the component being tested. These activation energies should be used in l all aging tests ::ubject t6 thE f 6116 wing celleria: I

                                                                                     )a Amendment 2                           4-20                                     I (June 1985)

{

(1) Tnermal pre-accident aging is to be done at a temperature v/ equal to or less than the maximum temperature the activa-tion energy was determined at. (2) Extrapolation of LOCA test data to determine post-accident operating time may be performed using the activation energy determined in a dry oven, as atmospheric condi-tions in the LOCA test chamber closely approximate those found in an actual post-LOCA Containment atmosphere. Using the above dry oven activation energy will result in a more conservative approach, since the high humidity environment of the LOCA test will exclude oxygen from the test chamber resulting in less thermal degradation. (3) The above extrapolation of LOCA test data to determine post-accident operating time should only be performed on the plateau portion of the temperature test curve that is equal to or less than the maximum temperature the activa-tion energy was determined at. The above criteria should be applied in all cases unless another approach can be justified. If the activation energy of an organic material used in the component cannot be determined, a value of 0.5 eV can be assumed. 4.5.3 MAINTENANCE AND SURVEILLANCE PROCRAM CONSIDERATIONS The above-discussed techniques have been used to establish periodic maintenance or replacement intervals for equipment with a calculated qualified life of <40 yr. The maintenance and surveillance program is described in Section 5.4 F 4.5.4 EVALUATION CRITERIA m The equipment is considered qualified for thermal aging effects if one j of the following conditions is met: 4-21 Amendment 3 (December 1985)

1

l J i

                                                                                      -I l                 (1) The equivalent thermal disgradation of the pre-aging exceeds the stated qualified life at normal temperatures for the area under consideration and the equivalent            l l

thermal degradation of the test profile is equal to or ! exceeds the equivalent thermal degradation of: (a) The specified temperature profile plus l (b) The assumed or required post-accident i operating time at maximum anticipated

temperatures for the most conservative l '

activation energy. (2) The equipment life is supported independently (eg, by l

  @                   operating experience / analysis or test) and the equivalent thermal degradation of the test profile is equal to or exceeds the equivalent thermal degradation of:

(a) The specified temperature profile plus l (b) The assumed or required post-accident operating l time at maximum anticipated temperatures for I the most conservative activation energy. I For this case, the accident test temperature profile may also be evaluated as a separate effects aging test. (3) A material analysis shows that the equipment is not i susceptible to thermal degradation (applicable to DOR j

i Guidelines only). j f (4) An analysis shows that the equipment of concern is similar to equipment that meets one of the preceding criteria.

A periodie replacement schedule must be established for equipment that has a qualified life of <40 yr. l Amendment 2 6 22 (June 1985) 1 1

.I

_ . . ~ q f l ! l v) 4.6 SUBMERGENCE 4.6.1 PIPE BREAKS INSIDE CONTAINMENT The maximum water level inside Containment resulting from a worst-case l large-break LOCA event is 52 ft 11.5 in. The bases used for determining this level are discussed in Section 15.6.5.5.2 of the Updated FSAR. All equipment required for post-LOCA/MSLB long-term operation either is l located above flood levels or is qualified for submergence. i Equipment required only for short-term post-LOCA/MSLB operation is automatically actuated by an Engineered Safety Features Actuation System j signal. Since the flood level calculation assumes the refueling water l storage tank, sodium hydroxide tank and all accumulator tanks are emptied 1 into Containment, Engineered Safety Features Actuation System actuation is prerequisite to significant Containment flooding. Therefore, the safety function of equipment in this category will be completed before flooding occurs. Flooding of this equipment after completion of its U safety function will not result in unacceptable failure of the flooded equipment, nor will it result in failure of other safety-related equipment that must be operable. 4.6.2 PIPE BREAKS OUTSIDE CONTAINMENT l No safety-related electrical equipment outside Containment has been identified which is prone to submergence following a HELB. Topical Report PGE-1004, " Trojan Nuclear Plant Analyses of Pipe System Breaks, Outside Containment", describes in detail the analyses performed, and specific features incorporated into Plant design, to assure that no safety-related equipment, electrical or otherwise, would be affected by flooding as a result of an HELB outside Containment. The effects of flooding outside containment from sources other than HELBs, vir, moderste energy line besaks, are analyzed and docuntented in _ e Trojan Updated FSAR Sections 9.2, 9.3, and 10.4. Certain protective D features, as described in the FSAR, are incorporated in Plant design to 4-23 Amendment 2 (June 1985)

preclude adverse flooding effects to safety-related equ*pment required b for safe shutdown or mitigation of the consequences of postulated accidents. 4.6.3 EVALUATION CRITERIA The equipment is considered qualified for submergence if one of the fol-lowing conditions is met: (1) The equipment type test included submerging the equipment and confirming its operability in a submerged state. (2) The equipment is located in a watertight enclosure which has been demonstrated to provide adequate protection by g test or analysis (the latter is applicable to DOR Guide-lines only). (3) The equipment is within the scope of the evaluation in Section 4.6.1 above. The submergence liquid shall be equivalent in composition to the liquid found in the containment Spray System. 4.7 NORMAL ENVIRONMENT ELECTRICAL VARIATIONS Normal environment electrical parameters that require c7nsideration are voltage, current, and frequency variations. 4.7.1 EVALUATION CRITERIA (1) Voltage variations are to be compared against values derived from calculations performed to determine the worst-case voltage expected during normal operation. These calculational results are contained in TE-032 en:1 show variations wititiu 210 percent for the 4.16-kV, 480-V a-c, 120-V a-c, and 125-V d-c h Amendraent 2 4-24 (June 1985)

systems, except for the 480-V battery chargers and inverters which vary within +6 percent.and -11 percent. (2) Generally, current variations are controlled by input ,; voltages or process inputs; therefore, explicit con-sideration of current variations is not generally required. Maximum currents should be considered when j 1 R heating effects are significant. (3) Frequency variations need not be considered signifi-cant as frequency control of the electric supply grid and emergency generators is very precise. 4.8 SYNERGISTIC EFFECTS synergistic effects result from the simultaneous action of discreet 13 ' effects such that the total effect is greater than the sum of the effects taken independently. It is eequired in 10 CFR 50.49 that synergistic effects be considered when these effects are believed to N have a significant effect on equipment performance. NUREG-0588 speci-fie9 that an investigation be performed to assure that no known syner- I gistic effects have been identified on materials that are included 8.n -! the equipment being qualified. The, DOR Guidelines do not require that synergistic effects be considered. L l Current qualification reviews performed for Trojan (excep', when l-reviewed to DOR Guidelines) seek to confit1n from the equipment vendor l- the identity of known synergistic effects on materials. Q'

4 4.9 MARGINS Margin is the difference between the most severe specified service conditions of the Plant and the conditions used in type testing to l account for normal variations in commercial production of equipment and reasonable errors in defining satisfactory performance.

lC 4-25 Amendmenc 2 ,e (June 1985) i k__-__-_______ - _ _ _ - _ - _ - - - . - . - _ _ - - - = _ - - .. - . _ - - . - . _ . - - _

l l

It is required by 10 CFR 50.49 that margins be applied to account for j g unquantified uncertainty, such as the effects of production variations l and inaccuracies in test instruments. These margins are to be in j l addition to any conservati::ms applied during the derivation of local environmental conditions of the equipment unless these conservatism l can be quantified and shown to contain appropriate margins. k l l g NUREG-0588 and Regulatory Guide 1.89 allow the following IEEE 323-1974 suggested values of margins to be used, in lieu of other proposed mar- I ! gins that may be found acceptable: i

                                        +15'F.

(1) Temperature: When qualification testing is l conducted under saturatec steam conditions, the I temperature margin chall be such that test pressure will not exceed satureted steam pressure corresponding to peak service temperature by more than 10 psi. 1 (2) Pressure: +10 percent of gauge, but not more than j 10 psi. (3) Radiation: +10 percent (on accident dose). ( *. ) Voltage: 110 percent of rated value unless other. rise specified. (5) Frequency: 15 percent of rated value unless otherwise specified. 1 (6) Time: +10 percent of the period of time the equipment is required to be operational following the DBE. l (7) Environmental Transients: The initial transient and

the dwell at peak temperature shall be applied at least i twice. I h

Amendment 2 (June 1955) ;

f (6) Vibration: +10 percent added to the acceleration of

   \~s'                the response spectrum at. the mounting point of the equipment.

These margins are incorporated into the environmental requirements for equipment qualified in accordance with NUREG-0588 and Regulatory Guide 1.89 (10 CFR 50.49). Additional margin beyond the specified ,, values is unnecessary. Margins need not be applied for evaluations I$ conducted against the DOR Guidelines. I l With respect to the 1-hr minimum operating time criterion imposed by the DOR Guidelines and NUREG-0588, the supplemental guidance of NRC Generic Letter 82-09 is used; ie, the 1-hr time margin rule is not applicable to equipment Whose safety function is performed prior to l significant changes in the environment at the equipment location, provided subsequent failures are shown not to be detrimental to Plant safety. g f'[ The post-accident operating time requirements given in Table 3-5 include the required margin, as well as post-accident temperature and pressure requirements given in Table 4-1. The post-accident radiation f$ values in Table 4-1 do not include the required margin. 4.10 TEST SEQUENCE It is necessary that the equipment type test simulate as closely as practicable the postulated accident environment. For equipment qualified by type testing, the following must be adhered to. 4.10.1 Equipment Qualified To DOR Guidelines The component being tested should be exposed to a steam / air environ-ment at elevated temperature and pressure in the sequence defined for its service conditions. Where radiation is a service condition which , is to be considered as part of a type test, it may be applied at any

      ) time during the test sequence or a separate effects evaluation may be f                                               4-27                      Amendment 2 (June 1985)

t l conducted, provided the component does not contain materials which are I known to be susceptible to significant radiation damage at the service condition levels or materials whose susceptibility to radiation damage is not known. If the component contains any such materials, the radiation dose should be applied prior to or concurrent with exposure to the elevated temperature and pressure steam / air environment. The ; same test specimen should be used throughout the test sequence for all service conditions the equipment is to be qualified for by type test-1 ing. The type test should only be considered valid for the service l conditions applied to the same test specimen in the appropriate sequence. 4.10.2 Equipment Oualified to NUREC-0588 Requirements NUREG-0588 requires that the guidelines given below, from IEEE 323-1974, be conformed tc fullyr (1) Inspection may be performed to assure that a test unit has not been damaged due to handling since manufacture and to determine basic dimensions. This inspection I shall not be directed to select a specific unit for j type testing. (2) The equipment shall be operated under normal conditions l 1 to provide a data base for comparison with performance under more highly stressed conditions. j (3) The equipment shall be operated to the extremes of all performance and electrical characteristics given in the l equipment specifications excluding DBE and post-DBE i conditions unless these data are available from other ; tests on identical or essentially similar equipment. ! l (4) Equipment shall be aged to put it in a condition which l l simulates its expected end-of-qualified-life condition including the effect of radiation. Certain key l l Amendment 2 4-28 (June 1985) I 1 !

                                   . -__                                                     ~

1 J 1 1 f- !

  - (f                                   measurements should be made following aging to deter-
                                                                                                                                  ]

mine if the equipment is performing satisfactorily prior to subsequent testing. (5) The aged equipment Phall be subjected to such meet n- q ical vibration as will be seen in service. This should I include simulated seismic vibration and self-induced vibration. (6) The aged equipment shall next be operated while exposed -I to the simulated DBE. Those functions which must be I performed during the simulated DBE shall be monitored. (7) The equipment shall then be operated while exposed to the simulated post-accident conditions (following i exposure to accident conditions). Those functions which must be performed following the simulated DBE ; ( shall be monitored during this simulation. (8) Disassemble, to the extent necessary for the inspection ; of the status and condition of the equipment, and record the findings.

   'l
   \                                                                      '

Amendment 2 4-29 (June 1985)

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I 5.0 ENVIRONMENTAL QUALIFICATION REVIEW (lLJ This chapter describes the overall approach used in the environmental qualification review and maintenance of electric equipment important lh to safety at the Trojan Nuclear Plant. lh l 5.1 REVIEW OF ELECTRICAL EOUIPMENT IN HARSH ENVIRONMENTS 5.1.1 REVIEW PROCESS The basic steps of the environmental qualification process used in the Division of Operating Reactors (DOR) Guidelines, NUREG-0588, and l 10 CFR 50.49, and followed by PGE in its review of environmental qualification of electrical equipment in harsh environments at Trojan, are as follows: (1) Identify all electric equipment important to safety. lb A V (2) Specify the environment equipment is exposed to during lh accid $nt conditions, eg, temperature, pressure, humid-ity, radiation, submergence and chemical effects. {b (3) Demonstrate the equipment will function in the speci- ih fled accident environment by one of the following methods: (a) Review testing an identical item of equipment together with supporting analysis to verify tret-ing demonstrates equipment's suitability for intended use. (b) Review testing of a similar item of equipment with f a supporting analysis to show that the equipment is acceptable. O ! k% l 5-1 Amendment 3 - December 1985 i
i (c) Review experience with identical or similar equip-ment under similar conditions with a supporting analysis to show that the equipment to be quali-fled is acceptable. (d) perform analysis using partial-type test data that supports the analytical assumptions and j conclusions. l (4) Establish and maintain a record of qualification docu-l mentation to permit verification that each item of electric equipment important to safety is qualified for O its application and meets its specified perforraance requirements under accident conditions. (5) Identify maintenance and surveillance requirements. These environmental qualification review process steps are illustrated in Figure 5-1. Chapter 3 of this report discusses the procedure used to identify, and provides ars identification of, the electrical equipment in the plant needed to function in an accident environment. Chapter 4 describes the environmental conditions and other parameters for which this equipment must be qualified, and the criteria used for evaluating equipment quali-fication for these conditions. This chapter describes the appenach and
^i g documents used in the environmental qualification review and mair.tenance process. Chapter 6 describes the documentation system established to verify the qualification of equipment important to safety.
~~5.1.1.1 Documents Used In The Review process The two major documents used in the electrical equipment environmental~~
~~]f O qualification review process are the Component Summary Sheet (CSS) and the Equipment Qualification Summary Sheet (EQS).~~
O' 1 Amendment 3 5-2 (December 1985)
j I ((o 5.1.1.1.1 Component Summary Sheet The CSS (PGE Drawing E-2) identifies those components important to th , safety that must function in a harsh environment and documents compli- l ance with environmental qualification requirements for those components. The CSS contains an individual sheet for each electrical component i "important to safety", as defined by 10 CFR 50.49(b). m 5 The CSS: (a) identifies each component by the Plant ID number, (b) identifies the component's safety function, service, and essential 13 performance requirements (c) identifies the postulated environment in which it must operate, (d) summarizes equipment data, (e) references ' the EQS and/or other qualification documentation, (f) summarizes the lh installation details, and (g) documents the electrical equipment environmental qualification conclusions. The latter three are comple-(, ted for equipment in harsh enviror.ments only and: * (1) Refersnce the qualification documentation which shows that the installed equipment will meet its functional requirements under normal and accident conditions. (2) Describe the installation in sufficient detail to allow verification that the qualification documentation applies to the installed configuration. 13 (3) Document assurance of component qualification for the installed application. Specific guidance on the preparation of the CSS is contained in References 11 and 12. O n 5-3 Amendment 2 t (June 1985) {
S.1.1.1.2 Equipment Qualification Summary Sheet The EQS is used to document and summarize review of qualification docu-ments. The intent of the EQS is to document the basis for certifying an equipment type for use in a given set of enviroranental conditions. The EQS also summarizes the important features of equipment qualification for b specific types of equipment in sufficient detail so that it may be used to conclude the suitability of the equipment reviewed for use in future applications without having to re-review the test report for every appli-cation. In order to minimize repeat reviews, the initial review is con-ducted against the requirements for the most severe environment for which qualification may be established, even if the current application is sub-ject to less severe environmental conditions. l l An EQS is prepared for each type of electrical equipment important to i safety that nest operate in a harsh environment. An EQS may also be pre-pared for an equipment type reviewed for suitability, but not purchased l or installed, or to document suitability of equipment installed in a mild j environment. In these cases, the completed EQS is still incorporated into the Qualification Central File for future reference. l i The qualification documentation may take the form of test data, operating I
)
experience data, analysis, or a combination of these. The latter two by themselves tre acceptable only for: l N I (1) Equipment evaluated for use in locations where radia-l tion is the or.ly harsh environment. g (2) Addressing the normal environment performance of equip-ment evaluated for use in harsh environments. i l i
~~@ (3) Supplementation of harsh environment testing.~~
b If after completing the EQS all requirements for the area in question j are shown to be enveloped by the qualification documentation or all g i deviations can be justified, the component is considered suitably l Amendment 2 5-4 (June 1985) t i
1 qualified for service in that environment and any environment where each environmental condition is less severe. Specific guidance for the review of qualification documents is con-tained in Reference 13. Ib 5.1.2 REVIEW PROCEDURE I l Review of the environmental qualification method of equipment impor- _1 ra J tant to safety involves two basic steps: 4 l (1) Review of equipment test reports to determine if quali-fication testing adequately demonstrates equipment operability under accident conditions. (2) Verification that Plant equipment is covered by the reviewed test reports. 3 V\ As discussed in the previous section, an EQS is used to document test , report reviews. These forms are included in the Qualification Central File together with the test reports as discussed in Chapter 6. A CSS is used to document the functional and environmental requirements of _ installed components and reference the appropriate EQS form. The procedures followed in reviewing test reports and their applicability to Plant equipment are discussed below for installed and new (including replacement) equipment. 5.1.2.1 Installed Eculpment All test reports in the possession of PGE when IE Bulletin 79-OlB was issued, and test reports subsequently received directly by PGE, were reviewed to establish the environmental conditions applied during the test. These conditions were initially tabulated on syster.and component [ p evaluation worksheets, and later on t.he CSS. 5-5 Amendment 2 (June 1985) i.
1 Additional qualification information was obtained from vendors and data ! review and analysis performed against the DOR Cuidelines (included as Enclosure 4 to IE Bulletin 79-01B). This review was documented by means ! of " problem files" for each equipment type. These problem files addressed each of the specific points discussed by the DOR Cuidelines. A written procedure was used for test report reviews and preparation of problem files. This procedure is included as part of the Qualification Central File. Walkdown and as-built data was also used in the review to ensute that the installed configuration of Plant equipment did not invalidate qualification. This review was documented as an appendix to each problem file. Subsequent to the reviews conducted in response to IE Bulletin 79-01B, g installed equipment was re-reviewed in accordance with the process
~
described in Section 5.1.1. This review was "backfitted" in order to assure consistency between the review approaches for existing and new equipment, j 5.1.2.2 New and Replacement Equipment G, i Qualification test reports for new or replacement equipment will be reviewed before installation. These evaluations will be performed in Cl accordance with References 14 and 15, as summarized below.
)
For previously reviewed equipment, the EQS will be checked to verify that the equipment is suitable for the intended application. The vendor will be requested to supply a certificate of Compliance, stating that the l qualification documentation in the file applies to the equipment supplied. For equipment not previously reviewed, a copy of the vendor's test report will be obtained and reviewed. If found acceptable, the test report will i be included in the Qualification Central File. The vendor will be requested to supply a Certificate of Compliance which states that the supplied equipment is of the same design and materials as the tested unit. I Amendment 2 5-6 (June 1985) 1
i j j Figure 5-2 provides an overview of the review procedure for qualifying ( new and replacement equipment. D 1 1 5.1.2.3 Nonconforming Equipment In cases where an engineering review identifies that a piece of equip-ment important to safety is not environmentally qualified in accor-dance with the criteria of Items I through 4.b of Section 2.2, a Non-conformance Report (NCR) is processed'in accordance with Nuclear Division Procedure (NDP) 600-1, " Control of Nonconforming Materials, ,) Parts, and Components" . NDP 600-1 establishes the measures for D i the identification, documentation, and control of nonconforming materials, parts, and components and provides for the documentation and verification of the resulting corrective action. Significant equipment qualification problems are evaluated for deportability in accordance with the requirements of 10 CFR 50.49(h) and 10 CFR 21 (via NDP 600-1). ( 5.2 REVIEW OF ELECTRICAL EQUIPMENT IN MILD ENVIRONMENTS Environmental qualification of electric equipment located in a mild environment (defined in Sections 3.3 and 4.0) is not included within 13 the scope of 10 CFR 50.49. In general, this means that electric equipment in mild environments are r.ot required to be environmentally qualified by type testing. Existing Plant equipment in mild environments is considered to be already tiualified on the basis of operating experience and current Plant procedures. This is because equipment failures occurring in normal environments have proved to be generally random in nature and are identified and corrected in accordance with existing Plant main-tenance, inspection and testing programs. Critical equipment (eg, in control room) is also protected against off-normal environments by redundant, emergency powered heating, ventilating and air conditioning systems. 5-7 Amendment 2 (June 1985)
a c
~~-.~.-.w.---.- .. .- . .- . . . . .~~
l For new Plant equipment, the requirements of 10 CFR 50, Appendices A and B are considered sufficient to ensure adequate performance of l 3! electric equipment important to safety located in a mild environment. l Design or purchase specifications will contain a description of the j l functional requirements and the specific environmental conditions g during normal and abnormal conditions. Existing Plant surveillance /
)
maintenance programs and periodic testing will also ensure, that such l equipment will perform its specified safety function during its 1 anticipated installed life. l 5.3 REVIEW OF OTHER EQUIPMENT SUP7t!i TO 1 i EtWIRONMENTAL QUALIFICATION i As discussed in Section 3.4, certain post-accident monitoring
)
instrumentation located in a harsh environment and installed in accordance with NUREG-0737 requirements was not initially subject to environmental qualification review. This instrumentation includes the i Post-Accident Sampling System, the noble gas effluent monitors, and the iodine and particulate sampling capability. The environmental i qualification requirements applicable to these instruments is summarized in Table 3-4. l The following approach has been taken for reviewing the environmental a qualification of the subject equipment: ' (1) Regulatory Guide 1.97 Category 3 design and , qualification criteria is applied to the subject equipment. That is, the equipment is demonstrated to I be of high-quality commercial grade and evaluated to { l determine that it can withstand the specified service I environment. I J (2) The service environments are specified on the basis of the accident environments documented in Chapter 4 (3) When considering radiation effects, an analysis is performed to: (a) identify radiation sensitive ! Amendment 2 5-8 (June 1985) u_________________
I materials, and (b) determine the susceptibility of these k'im materials to degradation. The guidelines specified in Sec-( tion 4.4.2 are used as evaluation criteria for this analysis. In general, radiation is the only environmental parameter of concern for the subject equipment. Therefore, material analyses are consi-dered an acceptable method, consistent with the DOR Guidelines and the intent of NUREG-0737 criteria, to demonstrate the environmental quali-fication of the subject equipment. S.4 MAINTENANCE AND SURVEILLANCE PROGRAM I Implicit in the NRC's environmental qualification rule (10 CFR 50.49) ! is the requirement for a maintenance and surveillance program which l' will maintain the qualification of installed equipment. Troj an's maintenance program has been modified to ensure that the following { requirements are met: 1 f (1) Maintenance which keeps electrical equipment in a qualified V condition. I (2) Surveillance which detects a known pattern of degradation. gl (3) periodic replacement which removes complete assemblies or subcomponents before the end of their useful or qualified lives and ensures the upgrading of complete assembly replacements to the newer requirements of 10 CFR 50.49, as implemented by Regulatory Guide 1.89. Revision 1 (unless there are sound reasons to the contrary, as specified in ! l Section 2.2). i With the uncertainties involved in the qualification process and in regulatory and technical issues, there is no attempt at this tim l extend the qualified life of electrical equipment. p 1 5-9 Amendment 3 (December 1985) j m
A detailed description of the maintenance and surveillance program is , h provided in the Trojan Nuclear Plant " Environmental Qualification Maintenance and Starveillance Program Manual" Reference 18, I i l I l l l O l l c l
~~) '~~
Amendment 3 5-10 (December 1985)
R E V A E T S I A M M A L L E L U Q N O P - LS U H N L M M S C O A RE O I T C ET T D T N - TS E N A E E SY E T C M T AS N A I S N A M O P FT O S DS DB Y P M I CT IN LE R T I E PT UE - E O I N AM VE E ZD C TE UE NH DH I E NM QR ES NS - RT E API I A - ER T EU EY Y TO A SU RQ RR ER - US R SQ AE AA SA P EE E PR MD S PM I M N M EM VM - ON E SO OO RU EU CA G AT CT PS RS
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~~..m._. ,c. , m m .~~
I 6.0 ENVIRONMENTAL QUALIFICATION DOCUMENTATION s.- This chapter describes the documentation system established to verify that each type of electric equipment important to safety is qualified for its application and meets its specified performance requirements. j i 6.1 CENTRAL FILE All documentation supporting the qualification of equipment important to safety in harsh environments is assembled and maintained in a Qualifica-tion Central File. This file satisfies the requirements of Trojan Technical Specification 6.13.2 and 10 CFR 50.49. i The Central File consists of Component Summary Sheets (CSS), a photograph file, and a records file of qualification and supporting documentation on ^l each type of electric equipment important to safety. The organization and content of the Central File is described in Reference 17. The 1 j Central File is maintained at PGE Corporate Headquarters in an auditable ) form by the Nuclear Plant Engineering Department. ' i 6.2 EQUIPMENT QUALIFICATION
~~SUMMARY~~
FORMS I Equipment Qualification Summary (EQS) forms are controlled documents used to document and summarize review of qualification documents such as: lb l (1) PGE evaluation of vendor equipment qualification test Ih ' reports and correspondence. I (2) PGE review of consultant evaluations of equipment Ib qualification. I i n I (3) PCE analysis, operating experience, or testing that l$ i establishes equipment qualification. i n j (4) Consultant evaluations of equipment qualification. Id
~~'u 6-1 Amendment 2 i (June 1985) i l l 1~~
l
hl The EQS forms document the basis for certifying a particular type of equipment for a specific application. Test report reviews are
~~@l governed by Nuclear Plant Engineering Procedure 200-21 . The completed EQS fccms are maintained in Section 4 of each record of the qualification Central File.~~
~~6.3 COMPONENT~~
~~SUMMARY~~
SHEETS 81 The CSS, also designated as PGE Drawing E-2, are controlled drawings used to document compliance with environmental qualification require-ments for individual items of electric equipment important to safety. The CSS: (1) Identify electric equipment important to safety by Plant ID number. (2) List each component's location, safety function, essen-tial performance requirements, and manufacturer's model number. (3) Describe the environmental conditions under which it must perform. In addition, for equipment that must operate under harsh environmental l conditions, CSS: l (1) Reference the qualification documentation which shows that the installed equipment will moet its functional requirements under normal and accident conditions. l (2) Describe the installation in sufficient detail to allow verification that the qualification documentation 1 bl applies to the installed configuration. (3) Document assurance of component qvslification for the , installed application, i l Amendment 2 6-2 (June 1985) l l I i l __ __a
Since this document is part of the Plant drawing system, it is included in the Central. File only by reference and-is maintained h separately from the qualification. records.
^
~~De preparation of CSS is governed by References 11 and 12. l O 6-3 Amendment 2~~
~~'(June 1985)~~
~~__ =__=__________________._______1 . _ _ _ _~~
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4
~~7.0 REFERENCES~~
((
~~% J)~~
1. Trojan Nuclear Plant Analyses of Pipe System Breaks. Outside Containment, PGE-1004, Revision 3, Portland General Electric Company (October 1975).

2. "A Nuclear Industry Position Paper on System Operating Times".
~~AIF Committee on Power Plant Design, Construction, and Operation (Subcommittee on Equipment Qualification), August 24, 1982.~~
3. Safety Evaluation Report Trojan Nuclear Plant, Docket No. 50-344 U. S. Atomic Energy Commission (October 1974).

4. Short-Term Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors, ;
~~NUREG-0458, D. S. Nuclear Regulatory Commission (May 1978). l~~
5. R. W. Braddy, M. M. Schoenhoff, J. W. Thiesing, " Surface l Temperature Response of Equipment Inside Containment Following Pipe Break", Transactions of the American Nuclear Society, 24 (1976).

6. M. Bruce, M. Davis, Radiation Effects On Organic Materials In Nuclear Plants, EPRI NP-2129, November 1981.
~~,s 7. F. Bouquet, J. Winslow, hadiation Data for Design / Qualification~~
[i ) of Nuclear Plant Equipment EPRI Research Project 1707-7, Final , l \hs/ Draft Supplemental Report, November 1984. t$ l 8. J. A. M rrell, et al, The Application of Remote Electronics In A Nuclear tuel Reprocessing Environment: Radiation Effects and Design Guidelines, SAND 82-2151, January 1983. I
9. S. P. Carfagno, R. J. Gibson, A Review of Equipment Aging Theory and Technology, EPRI NP-1558, September 1980.

10. Trojan Nuclear Plant Accident Monitoring Instrumentation Review, l PGE-1043, Portland General Electric Coupany (December 1984).

11. Nuclear Plant Engineering Electrical Branch Guideline 6 Electrical Equipment Environmental Qualification (Developmental Effort).
i^
~~12. Nuclear Plant Engineering Electrical Branch Guideline 7~~
~~. Component Summary Sheet (CSS). E-2 Drawing.~~
~~l{3 t~~
13. Nuclear Plant Engineering Procedure 200-21, Electrical Equipment Environmental Qualification Reviews. I l 14. S-509, Specification for Environmental and Seismic Qualification I of Electrical Equipment. First-Time Procurement.
O 7, Amendment 2 (June 1985)
l l
~~15. S-510 Specification for Environmental and Seismic Qualification of Electrical Equipment. Test Reports Previously Reviewed.~~
~~-- 16. Nuclear Divition Procedure 600-1, Control of Nonconforming 1 i~~
~~U Materials. Parts, and Components. j I~~
~~17. Record 0, Electrical Equipment Environmental Qualification i Central File Description.~~
~~-s 18. Trojan Nuclear Plant Environmental Qualification Maintenance and C Surveillance Program Manual. ,~~
I i i I i ! i l d l l l 1 I 1 l
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t ll I l l \ ; i I l l ! 1 ! l lh' Amendment 3 7-2 (December 1985) i L____---_-------- . I
8.0 APPE!OICES
+
This chapter ir.cludes appendices which supplement and support related 1 i information provided in the main text of this report. i > , I l l l l l l I l i I 1 8-1 i
~~-___. - - _ _ _ _ _ _ _ _ _ _ - - _ - - - _ _ _ _ _ _ _ _________-.______a~~
l i 1 1 8.A RADIATION DOSE ASSESSMENT 4 (h)N (,, This appendix' describes the Plant-specific analyses performed to establish the radiation specification values used for the environmental qualifica-tion review. Included herein are two component specific sample calcula- l tions (one for inside and one for outside Containment), and a description of'each of the methodologies used, their application and associated
]
I conservatism. i METHODOLOGY FOR POST-ACCIDENT DOSE ASSESSMENT l Total integrated doses (TIDs) for areas inside and outside Containment were evaluated for the LOCA and intact Reactor Coolant System (RCS) ! modes. Area dose rates for sources inside the Containment were deter-mined using the NUREG-0578 source term. Dose rate vs distance and time were calculated for pipes, vessels and other components containing reactor coolant. The ISOSHLD computer code, in conjunction with the RIBD fission product library file, was used to perform the analyses. l
~~\ INSIDE CONTAINMENT Loss-of-Coolant Accident (LOCA):~~
For the post-LOCA mode, the Containment Sprays,' Residual Heat Removal (RHR) and Safety Injection Systems were assumed to operate. Uncontami-nated water from the refueling water storage tank was added for the first 30 min, followed by recirculation from the Containment sump. The RCS was therefore diluted by the addition of the refueling water storage tank and accumulator volumes. The source term consisted of 50 percent of the core iodine inventory and 1 percent of other fission products uniformly mixed in the diluted RCS volume. One hundred percent of the noble gas inventory was assumed to have escaped into the Contain-ment atmosphere. e g l 8A-1 Amendment 1 (March 19?4)
l I i The dose rates were first determined for a 30 min decay period. Inte-grating factors were calculated based on the 30-min dose rate for time periods of 1 he, 1 day, I week, 30 days and 1 yr. These factors were obtained by performing ISOSHLD dose rate calculations for various time j periods, and then integrating the resulting dora rate vs time curve over ) l l ! the desired intervals. j S A spatial distribution of fission products in the Containment atmosphere was assumed, as outlined in Appendix D of NUREG-0588. This distribution model assumes that 100 percent of core noble gases, 50 percent of the halogens and 1 percent of other fiarion products are released into the Containment atmosphere, followed by Containment spray iodine removal. This results in the immediate removal of the 1 percent of solid fission products to the Containment sump. Initial fractions of the total core iodine released are assumed to consist of the following forms: 45.5-percent elemental, 2-porcent organic and 2.5-percent particulate. These values are in accordance with NUREC-0588 and Regulatory Guide 1.4. l Removal of iodine from the atmosphere to the sump is assumed to occur at i different rates and ultimate efficiencies for each of the three physical 1 forms. The ultimate fraction of total iodine remaining in the containment I atmosphere following 2 he of spray removal is assumed to consist of 0.24 percent as elemental, 1 percent as organic and 0 percent as particu-late iodines, with a total of 1.24-percent iodine retcaining. The noble gases released are unaffected by spraying. This distribution of source terms for the Containment sump and atmosphere required the calculation of separate integrating factors for each source. l In order to determine spatial variations and gamma doses for various compartments within the containment, dose correction factors based on Appendix D of NUREG-0568 were determined to account for the smaller volumes of the atmospheric source contributors within these compartments. Dose contributions from adjacent compartments, corrected by appropriate concrete chielding factors, were then added to obtain TIDs. O Amendment 1 BA-2 (March 1984)
~~.The end-of-cycle power history assumed for this evaluation consists of a . 2-yr full core irradiation at an 80-percent capacity factor. Operation~~
( was assumed at 100-percent power level for 80 percent of.a year, followed by 0-percent power level for 20 percent of a year and 100-percent power level for another 80 percent of a year. The RIBD fission product library was utilized to determine the source term nuclide inventory. Beta dose rates and TIDs for the LOCA mode were taken from tabulated values in NUREG-0588, corrected for differences between the Trojan reactor power level and Conta bment free volume, and then used as a basis for the NUREG-0588 values. Also, a contact beta source term was determined for components which are in direct contact with reactor coolant, for both the LOCA and the intact RCS mode. These dose rates are based upon a semi-infinite cloud approximation. For finite geometries which do not satisfy the infinite cloud conditions, as is the case for components which are enclosed but still exposed to the Containment atmosphere, volume correction factors were determined as a ( function of source radius (using ISOSHLD) and were applied, if necessary,
\
to determine beta dose rates'and TIDs within finite volumes. TIDs were determined by summing background and accident doses. Intact Reactor Coolant System: For the intact RCS mode, the primary coolant system was assumed isolated prior to severe core damage with core cooling provided by natural circu-lation and hydrogen desassing by reactor vessel head vent and pressurizer relief to the pressurizer relief tank. The RCS pressure boundary remained intact to add additional operator flexibility, however, the RCS letdown was assumed to operate until the holdup tanks were completely filled with coolant and the resulting gases vented to the Radioactive Gaseous Waste System. The RHR System was also assumed to operate. This results in a larger cross-hatched area (exposed to high Y-sources outside Containment) in Figures 8A-1 through 8A-10. 8A-3 Amendment 1 (March 1984) l
The RCS thus contained all of the noble gases with no dilution. The source term included 50 percent of the core iodino, 1 percent of the other fission products and 100 percent of the noble gases uniformly distributed in the undiluted RCS volume. Thirty-min dose rates from all sources were added together and then multiplied by the appropriate integrating factor to provide TIDs. piping runs, such as RCS loop piping, were sectioned into 10-ft lengths, and dose rate contributions from each 10-ft length were added to provide a total dose rate. Due to the low penetrating power of beta radiation, dose contributions , outside of piping or vessel walls were assumed to be negligible. J OUTSIDE CONTAINMENT LOCA/ Intact Y: i i Gamma dose rates outside Containment were determined by adding con-1 Sources tributions from all local sources for the location of interest. from piping containing reactor coolant were modeled as cylindrical sources.10 ft in length, of the appropriate diameter. ISOSHLD/RIBD was again used to calculate the dose rates vs distance from these 10-ft lengths. As for the inside Containment calculation, integrating factors based upon 30-min dose rates were determined in order to provide TIDs over a range of time intervals. The TIDs were again determined by summing background and accident doses. LOCA/ Intact 8: Beta dose rates were calculated for the LOCA and intact RCS cases for the inside surfaces of components contsining primary coolant. This was done using ISOSHLD to determine the beta dose rate caused by contact with the coolant. A set of beta integrating factors were then calculated in order to provide TIDs. Amendment 1 8A-4 (March 1984)
Due to the low penetrating power of beta radiation, dose contributions outside the piping or vessel walls were assumed to be negligible. .No (/ atmospheric cloud-type sources outride containment were considered. BACKGROUND D.1SE LEVELS l Background radiation dose levels were determined for 40 yr of operation. The background TIDs for safety-related equipment were taken from Section 3.11 of the original FSAR. l Table SA-1 presents the background Y levels ifor inside and outside l Containment. Separate background values for components located inside. the bioshield were determined in order te account for the significant dose contribution from N-16 in this area. OUTLINE OF METHODOLOGY Prior to the issuance of PGE-1025, electrical equipment qualification was performed using an essentially similar methodology. However, the methodology outlined in this report does differ significantly from the previous methodology in the determination of TIDs to areas outside Containment. The previous methodology used the following qualification criteria for components located outside the containment: the piping penetration area and all rooms containing piping with RCS liquids were considered to be areas directly exposed to high Y sources, equivalent to the cross-hatched areas in Figures 8A-1 thee ;gh 8A-10. Below Elevation 45 ft in the Auxiliary Building, the previous methodology used a qualification value of 4,5 x 10$ rads for all areas not directly exposed to high y sources, with the exception of hallways at Elevations 5 ft and 25 ft, where the doses were determined by specific evaluation to be
~~<1 x 10 rads. Above Elevation 45 ft, all areas not directly exposed to high Y sources were assumed to receive <1.0 x 10 rads.~~
8A-5 Amendment 1 (March 1984)
i The present methodology differs from this approach in that separate background and accident doses are determined for areas not directly I exposed to high Y sources. The background doses are determined by area 3 radiation zones (II through V), and a conservative value of 5.5 x 10 rads is added to these values to determine TIDs for all areas not directly
~~exposed to a high Y source.~~
l i 1 l The following methodology is used for conservative component TIDs for inside and outside Containment. When components cannot be purchased to TID values obtained using this methodology, a component-specific evalua-tion will be performed. Refer to flow charts in Figures 8A-13 and 8A-14 for appropriate actions corresponding to each step provide $ below: (1) Determine specific location of component (inside/ l outside containment, inside/outside bioshield, eleva-1 l tion, etc). l l ! (2) Detemine whether component is intended to function fol-lowing a LOCA or an intact RCS accident. Also determine l how long the equipment is intended to function following the accident, and the TID for which the component is qualified. 1 (3) Is the component in direct contact with RCS liquids? If so, determine contact Y and 8 doses from Table BA-1. ! l (4) If located outside Containment, is the component exposed i to a high Y source (cross-hatched areas in Figures 8A-1 ' l through BA-10)? = l l (5) Select appropriate Y dose from Table 8A-1, i I (6) If located inside Containment and the LOCA mode is being j considered, is the component exposed to the Containment I atmosphere? If so, determine S dose using Table BA-1 and Amendment 1 BA-6 (March 1984) l
Figures 8A-11 and 8A-12 (correct for finite volume if the exposed component is located within a confined space. [k,/ This is done using the methodology outlined in Figure 8A-12. (7) Select appropriate 40-yr background done from Table BA-1. (8) Determine TID by adding Y, g and background dose contributions. (9) Compare with TID for tdlich the component is qualified. If calculated TID is greater than rated, refer to Generation Licensing & Analysis for component-specific dose evaluation. EKAMPLE CALCULATIONS TNSIDE CONTAINMENT: Doses were determined for a valve operator motor on the RHR-line, component number M08701. (1) The components are located inside the Containment, at Elevation 55 ft 0 in., and outside the bioshield. )
)
(2) The components are intended to functien for 30 days following a LOCA. They are qualified to 2.0E8 rads. I (3) The components are not in direct contact with RCS liquid. (4) Not applicable. ! i i-l (5) Thirty-day Y dose = 2.7E7 R. (6) The areas containing electrical equipment have a confined volume <18,000 cm . This is equivalent to a source radius of 16.27 cm. Therefore, the 1-yr 6 dose is .26 of the infinite cloud dose or 3.6E7. l BA-7 Amendment 1
~~-(March 1984)~~
c i (7) Forty-yr background dose, outside bloshield = 1.0E7 R. (8) TID = 2.7E7 + 3.6E7 + 1.0E7 = 7.3E7 R. G l (9) Calculated TID less than rated TID, therefore the equipment meets qualification standards. l l OUTSIDE CONTAINMENT l s I l l Doses were detemined for the RHR pump motors. { (1) The components are located outside Containment, at , Elevation 5 ft 0 in, in the Auxiliary Building. ! (2) The components are intended to function for 90 days following an intact RCS-mode accident (for conservatism they will be evaluated for 1-yr operation). They are qualified to 2.0E8 rads. j l (3) The components are not in direct contact with RCS liquid. ei ' 1 (4) From Figure BA-6, the components are exposed to a high Y source (located within a cross-hatched area). 1 (5) One-yr y dose = 4.0E7 R. (6) Not applicable. l l (7) Forty-yr background dose, Zone 5 outside Containment = 4.5E5 R. (8) TID = (4.0E7 + 4.5ES)R = 4.045E7 R. (9) Calculated TID less than rated TID, therefore the equip-ment meets qualification standards. '- O Amendment 1 BA-8 (March 1984) E----_____________________________________________ _ _ _ _ _ _ _ _ _ _ -. - ------------- - - --- - - - - - - - - - _ _ _ _ __-.______u
T
~~B L E 8 * . 1 9~~
~~11/20/85 CALCULATED TSOJAN TOTAL INTEGRATED DORES~~
,. Y AND B (RADS) g ACCIDENT DOSE FIRST FIRST FIRST FIRST FIRST LEVELS _. INSIDE CONTAI g M s J Cg) g (g) g(g) MCOTHCg) g[g),
CA) LOCA. Y DOSE, BELOW EL 93 FT 1.tEG 7.1E6 2.0E7 2.7E7 3.3E7 (B) LCCA. Y DCSE. ABOVE EL 93 FT 1.7E6 S.SE6 1.0E7 1.3E7 1.3E7 (C) LCCA. B AIR DOSE. 8.1E6 4.6E7 9.6E7 1.4E8 1 *ES (INFINITE CLOUD)(B) (D) LOCA DOSE TO COMPONENTS IN CONTACT WITH REACTCR CCCLENTs CONTACT B DOSE 2.7ES 2.7EE 1.OE7 1.3E7 2.1E7 CONTACT Y DOSE 1.*E6 7.1EE 2.0E7 2.7E7 3.3E7 CD*) LOCA. Y DCSE TO SUBMERCED 1.CE6 1.0E7 3.3E7 *.6E7' 6.SE7 COMPONENTS i (E) INTACT R C S, Y DCSE. BELCW 7.7ES t, 7E 6 1.2E7 1.7E7 EL 93* 2.3E7 INTACT RCS. B AND Y DOSE N/A N/A N/A N/A N/A ABOVE EL 93' [F) INTACT RCS DCSE TO COMPCNENTS I IN CCNTACT WITH REACTOR CCCLENTs i CONTACT B DCSE 3.4E6 2.4E7 7. SET 9.9E7 1 '. E E 8 j CONTACT Y DOSE 4.7E6 2.9E7 7.6E7 1.CES 1.*E8 ACCIDENT DCSE ' ' LEVELS . DUT9IOE C CNT a T NMUNp (C) LCCA. Y DCSE. AREAS DIRECTLY 1.8ES 1.8E6 6.RE6 8.4E6 1.1E7 EXPOSED TO HICH SCURCE (CROSSMATCH 3(C) (H) LOCA DCSE 70 COMPCNENTS IN CCNTACT WITH REACTOR CCCLENT: CONTACT B DCSE 2.7ES 2.7E6 1.0E7 1.3E7 2.1E7 CONTACT Y DCSE 1 *EG 7.1E6 2.0E7 2.TE7 3.3E7 l (J) LOCA. Y DCSE. ALL OTHER AREAS (G) 7.EES 7.7EE 3.1E3 +.1E3 S.SE3 (K) INTACT R C S. Y DCSE. AREAS 1.SEE 9.0E6 1.3E7 3.2E7 4.4E7 DIRECTLY EXPCSED TO HICH SOURCE (. CCROSSHATCH)(C) (L) INTACT RCE DCSE 70 COMPONENTS IN CONTACT WITH REACTCR CCCLENT: CCNTACT B DCSE 3.4E6 2.4d7 7 SET 9.9E7 1. S E 8 CCNTACT Y DCSE +.7E6 2.9E7 7.EE7 1.0E8 1.+C8 [ M) INTACT RCS. Y DCSE. ALL DTHER 1.8E2 1.1E3 2.9E3 3.9E3 AREAS (D) S.SE3 enewcnCUND Deer LEVELST: 4D YR [N) Y BACMCRCUND. INSIDE CONTAINMENT: CUTSIDE BICSHIELD 1.0E7 INSIDE BIOSHIELD 2.2E7 (C) Y BACMCROUND. CUTSIDE CONTAINMENT ZCNE I S.0ER ZCNE II 1.1E3 i i ECNE III 6.6E3 ZCNE IV t,4Et ) ZONE V CAND II*, III*. Iv*) +.SE5 i 1 a (A) EQUIPMENT REQUIRED TO CPERATE 24TEGRATED DOSE (TID). 1 HR TO 1BETWEEN Opv USE01 HR SHOULD 1 DAY USE 1 DAY TC THE 1.HR 1 WEEK TCTAL USE i 1 WEEK. ETC. l l CB) SEE FIGURES SA.11 AND BA.12 FCR FINITE CLCUD B CORRECTION FACTCRS. (C) SEE FIGURES BA.1 THROUGH SA.10 FCR IDENTIFICATION CF CRCSSNATCHED AREAS CUTSIDE CONTAINMENT. ' (D) INCLUDES THE FOLLOWING AREAS THAT ARE NOT CROSENATCHED: AUXILIARY BUILDINC. FUEL BUILDING. MAIN STEAM SUPPCRT STRUCTURE. PIPING FACADE AREA. AND 4 ELECTRICAL PENETRATION AREA. FOR ALL REMAINING AREAS CUTSIDE CONTAINMENT. ' USE +E2 RAD POST. ACCIDENT DCSE REGARDLESS OF PCST. ACCIDENT EXPCSURE TIME. i Amendment 3 (December 1985) l J n no
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v t e Letter Sequence Other
TAC:64233, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML20210V194, ML20215N550, ML20235Z787, ML20236J406, ML20236N782
MONTHYEAR1986-09-25 [Table View]

ML20236N782

Text

1.0 INTRODUCTION

a W

=

SUMMARY

SUMMARY

7.0 REFERENCES

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