ML19220C078

Forwards SER Input for Facility Re Instrumentation & Control

ML19220C078
Person / Time
Site:	Crane
Issue date:	05/11/1976
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Deyoung R Office of Nuclear Reactor Regulation
References
NUDOCS 7904280182
Download: ML19220C078 (28)

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','n 1 1 G75 Dociet No. 50-320 j R. C. DeYoung, Assistant Director for Light Water Reactors, DPM METROPOLITAN EDISON CCMPANY, JEkSEY CEhTRAL PC'iER AND LIGiT COMPANY' AND PENNSYLVANIA ELECTRIC COMPANY, DIREE MILE ISLAND UNIT NO. 2; INSTRUME.VTATION, CChTROL, AND ELECTRICAL POiiER SYStiMS, SAFETY EVALUATION REPORT Plant Name:

Taree Mile Island, Unit No. 2 Licensing Stage:

Operating License Docket Su=ber:

50-320~

Milestone N.'mber:

24-22 Responsible Branch Lh12 and Project Leader:

H. Silver Systems Safety Review Branch Involved:

EIf.CS Branch Description of Review:

Safety Evaluation Report Requested Completion Dste:

April 20, 1976 Review Status:

Complete The enclosed Safety Evaluation Report (SER) was prepared by the DSS:PS, Electrical, Instrumentation and Control Systems Branch. This evaluation reflects the results of our review of the application through Amendment 40.

This report is submitted prior to = king a site visit, however, it does reflect the results of the electrical drawing audit.

Additional infor=ation, documentation, or resolution af concerr.s are necessary for the following (Reference SER Section):

1.

Acceptability of the present electrical, instrumentation, and centrol design provided for wiin steam and feedvater line isolation folicwing a imin steam line break accident (Section 7.3.2);

2.

Additional information and documentation concerning the Display Instrunentation necessary for Safe Shutdown anc Post-Accident Monitoring (Section 7.5);

3.

Co=mitment regarding qual.fication of the Hydrogen Recombiner to be used for this plant (Section 7.6.4);

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p,,3 y.~n+,1 n,,,14 m, un nc,,1,c +,,4 n, t, m.. of p1 ne Nuir-ent located outs Lde of the containment (Sechion 7.3.2);

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Co=mitment relating to the qualification program for Containment Electrical Penetrations (Section 7.9);

6.

Additional information and documentation concerning the design of the electrical controls for the Nuclear Services Closed Cooling n'ater Systen Pumps (Section 8.3.1).

This report identifies other items whirit require additional documentation.

The applicant has agreed to provide this documentation during the April 13, 1976 meeting.

031GIU G EI=

R. L. 03 Robert L. Tedesco3 Assistant Director for Plant Systecs Division of Systems Safety

Enclosure:

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'll:e Cor1:sion'; General Design Critaria (CLC), IEEE Scan _cas includia; IEE-2 Criteria for Protection Systms for .cclear Po.'2; Generating Stations (IEEE Std 279-1971 and IEEE Std 279-1953) and npp' teable Regulatory Guides for Pcwer Reactors nave been utili cd as tne bases for evaluating the adequacy of the prctection and control systcas. IEEE Std 279-1960 is the editica applicable to the Three Ele Island Unit Number 2 Reactor Prot 2ction Systca; however, th e U71 edition is applicable to the remainder of the design.

Specific c.curents employed in the review are listed in the Appendix to tc.:.s r eport.

The review of the protection and control systens was acconplished by ca.garing the designs with those of Rancho Seco and Three Mile Ia'and Unit Number 1 plants.

s Our review concentrated on those areas '..hich are unicue to Three Miic I31c.nd Unit Number 2, for which new information has been ceceis M, or whic'. have remained as continuing areas of concern dur;% this and prior reviews of similarly designed plants.

The en-ctrical drawing audit has been completed and this Safety E,,_1.ation Report (SER) reficcts the results of t'us audit and our b l O/ ve n' e" s

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7.2

'ctor_ Protection Systen (RPS) j T > design of the reactor protection systen is essantially identical to t::'.t of the Tnree Mile Isinnd Unit ; umber 1 plant. The systen utilizes the Babcock and ?lilecx Company relay logic design, uch w:-, first introduced for the Duke Power Cccpany Cconee Plants.

reactor prt tection systen is comprised of four redundant and independent reactor trip channels. Each channel incorporates ei.Nt bistables, any one of which, when actuated will in turn 4

ac: c.te a trip relay within the associated reactor trip module.

Sc e:t2d bistables of the eight in each channel represent on:

. cored parameter while the remaining histables represent con-in..

'o-.ution trips associated with more than one conitored param:ter infat.

The ci ht tri functions provided for each RPS channel are:

1.

hl;h r: ;;or coolant pressure; 2.

low react ; coolant pressur,

3.

hi;h reactor coolan: temper _:ure; B/

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reactor coolant pressure and temperature cor.carator; S.

neutron flux; 6.

reactor power and reactor coolant pump monitor comparator; 7.

reactor power imbalance; and, 8.

high reactor building pressure.

Each reactor trip module combines the four channel trip outputs in a two-out-of-four logic to trip the control rod power supply breakers. Each channel is testable during power operation.

The design has been previously reviewed by the staff generically and specifically for the Three Mile Island Unit Number 1 Operating License Application.

The applicant has identified one minor change involving the method used to derive the reactor power signal which is input (approp-riately isolated) to the Integrated Control System (ICS). Two of the four power range channels supply power signals to an averaging amplifie r and the remaining two power range channels supply signals to a separate averaging amplifier. The output signals of the averaging amplifiers are fed to an auctioneer device which selects the highest average power signal as an input signal to the ICS and power range recorder.

(For some previous designs this input signal to the ICS is provided by a single pcwer range channel as selected by the operator). This change does not violate any of the applicable portions of IEEE Std 279-1968 and is acceptable.

87 186

4 7.2 The applicant referenced SAW-10003 Topical Report for the RPS equipment qualification testing. The staff has recently completed its generic review of this document noting that the applicant must demonstrate that the equipment will comply to certain interface criteria in its as installed condition. The applicant has agreed to document this information with justification where exceptions to the interface criteria are taken. We conclude that this commitment is acceptable pending our receiving this documentation formally. We will report the results of our review of this documentation in a supplement to this. report.

7.2.1 Anticipated Transients Without Scram (ATWS)

The Regulatory staff's requirements with respect to ATWS are provided in the staff's technical report on " Anticipated Transients Without Scram for Water-Cooled Power-Reactors", WASH-1270, dated September, 1973.

The programs for implementation of these provisions in the Three hule Island Unit Number 2 Plant are provided in Appendix A, paragraph II.B of NASH-1270.

With regard to ATNS, the applicant has referenced the Babcock and Wilcox topical report, BAW-10099, " Babcock and Wilcox Analysis of Anticipated Transient Without Scram." The applicant has advised the staff (by letter from R. C. Arnold to A. Giambusso, dated February 14, 19/5) that on the basis of this repor they have concluded that neither design modifications to the Three Mile Island Unit Nurber 2 plant nor FSAR revisions will be required in order to conform to the requirements of NASH-1270.

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S 7.2.1 The Regulatory staff has completed its review of BAW-10099 on a generic basis and xus formulated a position pertaining to implementa-ton of ATWS requirements for Babcock and Wilcox plants. This posi-tion does not explicitly identify instrumentation, controls or electrical power system design changes for these plants. However, should future developments explicitly identify any design change in these areas we will require Three Mile Island Unit Number 2 to conform to these design changes.

7.3 Safety Features Actuation and Engineered Safety Features Systems The Safety Features Actuation System (SFAS) is comprised of para-meter monitoring channels, basic logic channels and two independent and redundant component actuating trains. The monitoring channels consist of three independent and redundant sensors (with associated circuitry) which monitor reactor coolant system pressure and reactor building. pressure. Each channel is testable during reactor operation.

Two sets of monitoring channels are provided to monitor the reactor building pressure with different actuation setpoints for each set, thus permitting different equipment to be acutated at each setpoint.

Each monitoring channel is coupled to three independent and redundant logic channels which actuate (on a two-out-of-three basis), two redundant Engineered Safety Features (ESP) component trains.

Our review of the Three Mile Island Unit Number 2 application in-cluded the review of selected detailed schematic drawings of the cir-cuitry pertaining to the SFAS and also the circuitry pertaining to n

the individual components; e.g., pumps and valves of the ESF systems.

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7.3.1 S?AS Differrces - Three ':ile Island Unit :; umber 2 vs.

G. 2;. in e Island unit

..:scr 1 The SFAS for the Three Mile Island Unit ';unber 2 plant is functional 1 id2ntical to the one for the Three :.lile Island Unit Number 1 plant; however, they differ in the following respects:

1.

For the Unit Number 2 plant, high and lo.. pressure injection are initiated at the same low reactor coolant systea pressure set-point. This is possible because the decay heat removal syste:a d'ich is used for low pressure injection is designed to cperate continuously against a closed head. For the Unit Number 1 desi;n hi;h a_.d low pressure injection are initiated at different values of the reactor coolant system pressure.

2.

Unit Number 2 uses pressure switches instead of pressure trans-mitters to monitor the reactor building pressure.

he have reviewed the above changes and have concluded that the resulting designs equal or e.xceeds those of the Three :lile Island Uni.: Nunber 1 plant, meet the requirements of IEEE Std 279-1971, and a :- acceptable.

7.3.2

~ id::".ter 2nd Stean Line Isolation 1 a cussion of the main steam line break refer to Sections C.: and 12.

of this report.

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trements. Hence, we will require additional iaformation from, the

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a..2 tion con #orm to the above requirements. ;ie will report the resolution concernine this it:m in a su'alement to this raport.

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rcire this area. Should it be determined

• hat this isolation u e:. :.t.2 for acceptable consequences followin; a steam lina break at..;ent, we will require that the electrical, instrumentati n, and ce :ro. prtions necessary to isolate the steam system to satisfy the s.c.;1e failure criterion. This would also require that the rei tie.nt lo;ic circuitries be routed in accordance with the phys:. cal se prati..- :;iteria for safety systems. de will report additional re s _Its f:r th.3 item in a supplement to this SER.

7,4 Sys e.s bcuired for Safe Shutdown T.m sufa s:- ::down systems as identified in the FS.\\R are:

Controi Rod Drt',e Control, Makeup Pump Control, Letdown Lina Isolation Valva

8 7.4 Control, Borated Water Storage Tank (SNST) Suction Valve Control, Emergency Feedwater Control, Pressuri:er Control, Decay Heat Removal System Control and supporting systems necessary for the proper operation of these systems.

Our review of the Three Mile Island Unit Number 2 application in-cluded the review of selected detailed schematic drawings of the cir-cuitry pertaining to the systems required for safe shutdown, including the circuitry used to initiate operation of individual components (e.g., pumps and valves). Significant items resulting from this review are discussed below.

7.4.1 Emergency Feedwater System (EFS)

The EFS consists of one turbine driven pump, two motor driven pumps and associated piping and valves. Diversity of control power is provided, such that, in the absence of all a-c power the turbine driven subsystem could operate. Actuation of the system occurs on loss of main feedwater pumps, loss of all four reactor coolant pumps, loss of power or manual operation.

The original design as proposed was modified because in the event.

of a steam line break accident total reliance for auxiliary feed-water flow to the unaffected steam generator was dependent on the operation of the Integrated Control System (ICS) and the Air Supply System which are non-safety grade systems. This design was sub-raquently modified to include bypass valves around the diaphragm operated valves which are controlled by the ICS and the iir Supply System. This provides an alternate parh for feedwater flow which is S~l iG.1

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,calicabic criteria and are acce1ntable.

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kuiliary Contral Stations E;

'"t:.11 instr =entation and controls are prast.ced cutside of tha

.: air. contrcl roca far the purpose of achieving a shutCri. condi.icn Th ;s 2 centrels and inst:=ents arc adecuate to achieve and acintair t.:. ait in a hot shutdoun condition and aise provida capability fer at:?.inin" a cold shutdown condition when suppler.en ced bv an. crc.ariate c7eratar actions. ?le have concluded that this des un cats the ap,'li-cu le criteria and is acceptable.

7.4.3 Decay Heat Rcroval (DHM Low Pressure to Hith Pressure Isolatica L 1,as T.:e ator-operated valves used to prevent o capras;t.rization o f the DS -, sten by the Reactar Coolant Syst:n ar requi..l to ccafo= to the inilcuin; criteria:

1.

V-a nives in series to isolate the PHP low pressure systca from the high pressure syste..

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Tha motor-opcrated valves are closed automatically wha uver the reactor ecolant syst:.a pressure exc:els the pr3csure ratin; cf tha lo.. pressure system The closure decices are desi,oned to comply '..ith the requirements of IEE2 Std 279-1971.

4.

Suitabla valva position indication should be provided fc r thase valves in tha control roca.

1'c have reviewed the detailed d:si;r. of these motor-operated valvas for conformance to the above criteria. rie conclude that the elec-trie:1, instrumentation and control aspects of their design s1t'.sfy

.ese criteria and are acceptable.

7:

Saf 2ty-Related Display Instrumentation (SI'01)

The safety-related display instrumentation provides information :a the reactor operator to enable him to perform required saf2ty func-tiens duric.3 normal operations, abnormal operational occurrences, ans accident and post-accident conditions.

Our review of the SRDI included the features for monitorin; of the Ei systeru, ESF support systems, safe shutdown systems and accident and post-accident cond1tions. We have not:d c.uring tne course or tc;s

11 7.S review that the disp;ay instramentation 'equired for. safe shutdown and post-accident aonitoring has not be en seismically qualified.

For tne above functions, the spplicant has agreed to explicitly identify the critical parameters to be monitored. Also, for these par meters, additional cesign information will be provided defining a

the extent to which the instrumentation for these parameters conform to safety criteria (i.e., seisuic qualification, physical separation, redundancy, etc.).

he wi.'.1 review this clditional design information and will report the resolution for this item in a supplement to this report.

7.6 Other Instrumentation Systems and Additional Requirements Recuired for Safety 7.6.1 Changeover from the Iniection Mode to the Recirculation Mode of Operation Following a Loss-of-Coolant Accident The original design as proposed accomplished the changeover from the injectier. mode to the recirculation mode of operation following a loss-of-coolant accident completely manually. Concern was expressed nbout the number of operations which the operator would have to complete, and the time available during which these operations must be correctly completed so as to not to degrade the emergency core cooling system (ECCS) pump performance due to loss of suction head.

As a result of these concerns the applicant has documented that this feature of the design will be automated. Also, this automatic feature is to be initiated on low Borated Water Storage Tank level.

In addition, the applicant has committed to provide the revised elec-trical schematics and diagrams :hich show how this automatic feature 4

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Valve position visual indication (open or closed) for each valva tnat is not de.cendent on oo'.;cr bein availabi' to the valte

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...r. dant visual and au ible alarms based en isclation ' tive psit:. :n a re p rovic..1.

The alarms are actuated if either val.e is ciascd while react >r coolant pre.uure is above the alar.

s e o,.n.: level. These alares are actuated by redundant.ad ind; ;.. dent valve position. sensing circuitry, includin; en:

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13 7.6.2 and independent pressure signals. Another alarm is actuated if either valve is open and reactor coolant pressure is reduced to a value that could cause emptying of the core flooding tanks; this alarm is intended to alert the operator to an impending situation where inadvertent discharge of the core flooding tanks could occur.

3.

The breakers that supply power to the motor operators for the isolation valves will be locked open and so tagged after the valves are opened. The Technical Specifications require this action as a prerequisite for bringing the reactor to criti-cality.

We have reviewed these features and conclude that they are acceptable.

7.6.3 Bypassed s"<'. Inoperable Status Indication for Safety-Related Systems Our review for this area concentrated on the prcvisions included in the design to indicate to the control room operator the inoperable status of the safety systems or subsystems. The systems included in this review were the RPS, SFAS, ESF systems and vital support systems for the ESF. The applicant has documented that the unit has been designed so that if a safety system or subsystem is deliberately rendered inoperable, one or more of the following occurs.

1.

Inoperable status is annunciated in the control room; 2.

For engineered safety features, the bypass is automatically overridden when tnese systems are required to operate; and, 87 196

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The bypass can be overridden by the control roca operator for the safety-related sys tems which support, but are not connected to, engineered safety feature systems.

Exceptions to the above are the felicwing components which have local controls that are not alarmed or indicated, and cannot be overridden from the centrol room:

a.

Reactor Coolant Make-Up Pumps

b.. Reactor Building Emergency Cooling Water Booster Pumps, lnd c.

Control Building River Water Bactter Pumps.

The local switches associated with these cosponents are key locked and strict administrative controls are enforced for these keys to prevent inadvertent disabling of the systens. Also, whenever any cf these components are rendered inoperable, appropriate back-un com-ponents are actuated, or placed in a suandby status. We will require that Three hule Island Unit Number 2 Technical Specifications shoul d appropriately include the above referred eperating procedure for these components.

We will report the resolution of this item in a supplement to this report.

7.6.4 Ccebustible Gas Centrol System (CGCS)

The CGCS consists of the containment hydrogen recombiner system and as a backup to this system the containment hydrogen purge system.

The applicant has documented that the Atomics International thermal, 87 197

15 7.6.4 external hydrogen recombiner will be used at the Three Mile Island Unit Number 2 plant. The applicant has also referenced the Atomics International Topical Report, AI-75-2, " Thermal Recombiner Systems for Water-Cooled Reactors." The staff has concluded from its review of this topical report that a seismic type test of the recombiner unit is necessary for qualifications. Atomics International has agreed to perform such testing.

We will therefore request that the applicant make the commitment to correct any deficiencies revealed as a result of this additional testing. We will ~ report the results of this item in a supplement to this report.

7.6.5 Protection System Response Time Testing The application has provided capability for testing response times of the protection systems.

In addition, t'ae applicant has documented that the Three Mile Island Unit Number 2 Technical Specifications will be revised to co=mit to performing response time testing of the Reactor Protection System and the Engineered Safety Features Actuation System sensors as well. Also, a response time test program will be submitted to the staff for review prior to implementing such a test program during Three Mile Island Unit Number 2's first refueling outage.

We conclude the this commitment is acceptable for protection system response time testing.

87 198

16 7.7 Control Systems Not Recuired for Safety The following control systems are identified in the FSAR as not required for safety:

the Unit Control Systems, the Integrated Con-trol System (ICS) and the Control Rod Drive Control System. The FSAR further separates each of these control systems into subsystems.

We have reviewed these subsystems for each control system and con-clude that their design is similar to those of other licensed plants.

The applicant has stated throughout the FSAR that the safety analyses assumes no credit for any ICS function which might be available to prevent or mitigate the consequences of an accident. We interpret this to mean that the safety analyses were conducted assuming the worst credible failure of the ICS in the active state in addition to its failure in a partial or totally inactive state. Accordingly, we have concluded that the differences in the desipi details are minor, do not affect our previous conclusions and that the design of these systems are acceptable for the Three Mile Island Unit Number 2 plant.

7.8 Qualification of Safety-Related Electrical Ecuir. ment 7.3.1 Seismic Cualification The applicant has stated that the Category I electrical equipment has been seismically qualified by prototype tests or analyses in accordance with IEEE Std 344-1971, " Seismic Qualification for Class 1 Electric Equipment for Nuclear Power Generating Stations." In addition, the applicant has identified this equipment, described the seismic qualification programs, summari:ed the test and/or analysis results, and identified the test documentation.

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.aa' ifi;2:ica Th: 2 y:.icant has identified in the F3.G all instrumentatic.:, control and clectrical equipr.ent important to saft.,..

Further, it has identified the location of this equipa:n and stated the expected limiting environmental conditions at that locatian. Additional;;,

it has stated that qualification tests and analyscs have demonstrated the ability of the safaty-related equipment to function in the normal and post-accident environment in accordance with these expected li...itin; enfironmental conditions. However, recent re-analyses (See Sectia.t o.0 ef this report) performed for this plant have indicated that the contain-cent vapor temperature exceeds briefly that of the containment desi;n bases tecperature for a short period of time following selected steam line breaks within containment. The applicant has documented that an analysis has been performed '.hich shows that this vapor temperature does not result in any safety-related equipment temperatura in excess of the temperature for which the equipment has been qualified..\\dditional'y, the applicant has agreed to provide additional documentation to support this conclusion.

We have also notei during the course of our review that sel2cted Class 12 electrical equipment within the balance of plant scope of supply '

3 no sp-cial environmental conditions specified in their e'.q..ent design ;pecification requirements. Equipment within this (g) 6/

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'-:: til.' review this additional information and rersr: the r2sul:1 of our review of thase environmental qualifica i.- cast; aad their results (for the balance of plant Class IE systa. c : 7onents) in a su:plement to this report.

7.9 C taintant E1:ctrical Penetrations Tc' a.olicant has stated that all cont 1innent electrical..ane: rations z

caet scismic CJ us I requirements and are in accordanca with the r'. _irements of taa applicable ASS'E code. Addition lly, the cca-ta ament electr. cal penetrations are designud to mae: all the elac-tr_ cal requirements for the service environment without dielectri:

b ra.<down or overheating and are qualified for tha sortico envir-o r.r.ent.

T he rp'.ican his further stated that the containaant electr cal

a ations 225cnblies meet the requirements of IEEE S:d 317-10 :,

E '. e c ital Penetration Assc blies in ContaiMant Structures ic 8~ 20!

19 7.9 Nuclear Fueled Power Generating Stations." Also, a description of the prototype test as performed by General Electric and the results of this test are contained in the test report-GE Number 74-502-3.

Additionally, the applicant has documented that both the electrical and environmental test conditions as noted in this test report ex-ceed the design electrical and environmental conditions for Three Mile Island Unit Number 2.

We.' ate, however, that the GE test report has not been reviewed by the Regulatory staff. Therefore, we will require a cocnitment fron the applicant to adopt any generic resolution of the staff's concerns resulting from the Regulatory staff's review of this test report for the Three Mile Island Unit Number 2.

We will report the resolution on this item in a supplement to this report.

7.10 Fire Stops and Seals The applicant has responded to our request concerning the design criteria and procedures for the fire stops and seals to be used at Three >dle Island Unit Number 2.

We have reviewed this information and conclude that it provides an adequate response on this subject.

We note that this general subject including criteria for ire de-tection and fire protection systems are being currently reviewed by the Regulatory staff. We will require Three Mile Island Unit Sun-ber 2 to conform to any additional requirements emerging out of the staff's generic study on this subject.

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20 3.0 ELECTRIC POWER S.1 General The Commission's General Design Criteria 17 and 1S, Regulatory Guides 1.6, 1.9, and IEEE Std 308-1971 were utilized as the primary bases for evaluating the adequacy of the electric pcwcr systems of the Three Sele Island Unit Number 2 plant. Specific documents used in the review are listed in the Appendix to this report.

8.2 Offsite Power Systems The substation at the site is a 230 kV breaker-and-a-half switching arrangement which provides terminal facilities for three trans-mission circuits and the unit auxiliary transformers. Two of three transnission circuits go north to Middletown Junction on different double circuit towers to connect the substation to the existing Metropolitan Edison Company (SEC) 230 kV transmission network. The third circuit connects the substation (on a different right-of-way) to the existing SEC 230 kV transmission network at Jackson. The two rights-of-way are divergent. This arrangement ensures that at least one circuit is available if structural collapse occurs within a right-o f-way.

Primary and backup protective relaying systems have been provided for each 230 kV circuit in addition to circuit breaker failure protection.

Two redundant and separate sources of d-c centrol power are supplied to the 230 kV substation from Three Mile Island Unit Number 1 250/125 volt d-c sysces.

Loss of either d-c source should not inhibit the ability to supply offsite power to tae station.

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21 S.2 Offsite power is normally supplied from the 220 kV substation to the two unit auxiliary transformers by two physically independent circuits.

This arrange cat provides for two immediate access sources. Each of the two unit auxiliary transformers (230/4.16 kV) normally con-nects to one of the two 4.16 kV ESF buses. The two 4.16 kV ESF buses are arranged in a two division split-ous configu'ation.

In the event of a failure to either circuit, transfer of the loads to the remaining source is accomplished automatically by relay and breaker action. Each unit auxiliary transformer is sized to carry the unit full load auxiliaries and the engineered safety feature auxiliaries.

Steady state and transient stability analysis have been made to determine the performance of Three Mile Island Unit Number 2 as well as the transraission network during contingency situations. The results of these studies have shown that no unit instability, system instability, transmission line overload or cascading outages will occur as a result of a 3-phase fault and ou* age of any transmission line emanating from either the Three Mile Island 230 kV or 500 kV bus. A;. itional studies have shown that the sudden loss of the catput of Three Mile Island Unit Number 2 by itself or along with the next largest unit (Three Mile Island Unit Number will not result in any system or unit instability, transmission overloads, cascading outages or intolerable voltage conditions in the network. Also, the planned transmission system meets the Mid-Atlantic Area Council's

" Reliability Principles and Standards" and has been approved by the Council.

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22 8.2 We conclude that the design of the offsite emergency power system-satisfies GDC 17 and 13 and is acceptable conditioned only on verification of the design installation by a site visit.

8.3 Cnsite Power Systers 8.3.1 A-C Power System Onsite standby power is supplied by twc 3000 kW diesel generators.

Each diesel supplies one 4.16 kV emergency bus which is associated with one division of the two-division split-bus configuration.

Inter-locks are provided to prevent parallelir the diesel generators.

Each diesel is automatically started by an undervoltage signal from its respective bus or by a safety feature actuation signal. Only one of the two diesels is recuired to provide emergency power for accident conditions.

The redundant engineered safety features and vital instrumentation and control loads are supplied, directly or indirectly, from the two 4.16 kV emergency buses through the two-division split bus con-figuration. This configuration is maintained throughout the a-c and d-c subsystems.

Interlocks are provided to prevent redundant buses from being paralleled by tie breakers.

However, we have identified features of the design enabling a single component to be powered from either of two redundant ESF sources.

The applicant has stated that interlocks are provided to prevent 8/

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23 8.3.1 these components from being powered from more than one redundant source simultaneously. We have verified that these interlock features have been included in the design during our electrical drawing audit. However, during this aspect of the review we also noted that for the Nuclear Services Closed Cooling Water System (NSCCWS) pumps, the existence of a feature which automatically swings a third pump between two redundant ESF 480 volt Motor Control Centers (MCC's). This automatic feature is presently retained for both normal and accident conditions. Our concern with this automatic feature is that it may compromise the indepen-dence of two redundant ESF 480 volt MCC's.

During the meeting held on April 13, 1976, the applicant stated that this design has been revised to preclude this automatic feature if accident conditions are present. Also, the applicant has agreed to provide adequate bases for retaining this automatic feature for normal plant operation. We will review this additional information including the revised electrical schematics and will report the results of our review on this item in a supplement to this report.

The diesel generators are rated at 3000 kW continuous and 3300 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The maximum emergency load they will be required to carry is 2416 kN for a period of 40 minutes. This is well within the recommendations of Regulatory Guide 1.9.

The applicant has also stated that the diesel generator units being furnished to Three Mile Island Unit Number 2 are identical to those supplied to Peach Bottom 87 206

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uaits is :'on ceemed nacassary Each diesel generator and its au:aliary syst:a is separately cacioseu in a seismic category installation.

The starting and operation of any diesci is not candition

'y operation of the other. Each diesel will receive a starti..

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' hen any of the following occurs:

1.

loss of essential bus voltage, 2.

a safety feature actuation signal, 3.

anual start (local or remote), and, 4.

test signals to simulate any of the above.

Each energency generator is equipped with mechanical and electrica' trip interlocks to ensure personnel protection and to pr: vent or licit equipment damage. Our position with regard to diesel generatar trip devi:es is that these devices (with the exceptica of en,,ine overspeed and generator differential) should be bypassad if a safety features actuation signal is present and the bypass circuitry should comp',e with the safety criteria or that the trip function be formed wit'.

cincidence logic and be completely testable. The documentation wh;;h the applicant has provided clearly indicates ::::: tha desi;n o7 7P7 bI Ld;

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6 W. h ve also noted during our rr'iew that the un 'ervolta;2 relay-at the 4.16 kV ESF buscs were not peria.acally tastable duri.;

r cal powar operations. During the meetin3 held ca Ap;.il 13, 10 3, the applicant agreed to provide a capability co test thesa under-voltage relays periodically and to submit revised cloctrical sche.r.atics.

W conclude that this coraitaent is acceptable and vill report the results of our review of the codified schematic concerning this item ir a su,plenent to this report.

E ch dicsci engine is supplied with a 25,000 ;211cn diesel fuel tank ar.1 a 350 gallon day tank with two 10 gallon per.ainute fuel oil t.;'.sfer p. ps which automatically maintain the level in the day t a;. 's. Sufficient fuel is stored to allow one unit to suu..nl':.aost-accident power requirements for seven days.

Th2 stativa 120 VAC system consists of five channels of 120 Velt

,.-c vital po ar each supplied nor ally via a static suitch frca a 250'. dc/

120 VAC inverter, or as a backup a 430/120 VAC re;ulated ' col:a;e each sup-

._sc from one 4SO/120 VAC regulated transformer. We w', 2 noted : hat for the system level Safety Features Actuation Signal the:. exist a n

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r:p e t the results of cur reviou of this Ja; centati'n in a su.a.aleuant to this report.

With the exception of the aforamentioned itera, t.c concluda -l'at the deni;n of the.: ergency a-c power systa nee:s GC 17 and 13.

IEEE Std 308-1971, and Re.mlatory Guules 1.6 and 1.9 :n:1 is acce.n-table; conditioned only on the verification of the design installatier by a site visit.

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' Pcuer Systen T:.: d-c pauer sys cm for this unit consists of tuo independant 250/125 voit d-c subsystans. Each subsysta:a utili:es three isolat d phas; buses and is supplied by a. battery and rectifiers.

The batteries are rated at 1800 rpere - hours on an 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> dis-charge basis. Following the loss of a-c power, the batterles are sized to supply all d-c safety loads for a period of tuo hours.

Each of the four rectifiers provided for each battery are rated at 45' VAC, 60 hert: input with 125 volts, 400 naperes cutput. These out ats are connected in serias narallel to fo = the 250/123 Vol:

d-c, 3 uire isolated bus sections.

Each of the units two sa:s :I re:-ifiers are capable of supp1 vin" the normal contimcus cirect o

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27 3.3.2 current load of the unit while recharging the battery to a fully charged state from a discharged condition within 4 to S hours.

The components of the d-c power system are located in the Control Building, a Seismic Class I structure. Additionally, the d-c buses and the associated equipment are physically located such that the redundant counterparts are separated from each other.

The vital d-c system is compatible with the two-division split-bus configuration of the a-c system.

We have concluded that the design of the vital d-c power system satisfies the Co=nission's requirements and is acceptable conditioned s

only on verification of the design installation by a site visit.

8.4 Separation and Identification of Safety-Related Ecuipment and Systems The applicant has documented in Sections 3.3.1.4 and 3.3.1.S of the FSAR the separation and identification provisions included in the design. The applicant has also documented the degree of conformance with Regulatory Guide (RG) 1.7S.

With the exception of the specific separation distances recommended in this guide the applicant has stated that the design of this unit meets the intent of RG 1.7S.

We will audit selected areas where less stringent separation distances are used during the site visit. We will report concerns, if any, on this item in a supplecent to this report subsequent to the site visit.

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72-1 a:..d Cca t rol Sys tc.2: 5:ench in tb; r2 vie.; of the..: r :e

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Appendi.. A to 10 CFE Part 30, datel July 1971 3.

Regulatory Guida.4 1.6, 1.7, 1.9, 1.11, 1.22, 1.32, 1.40, l.,1, 1.47, 1.53 and 1.75.

4.

Three '21e Island Unit 2 ::uclear Generating S:2 tion Final Safe:y Analysis Report (FSA;O through Amendrant 40.

5.

Safety Evaluation Report - Ranch Seco Nuclear Cen2r2 it.; Stacint.

Unit 1, da:ad June 3,1973, Davis Sesse s:aclear Pc.ter S ta: ion Uait 2, dated June 23, 1975.

6.

T.:e followin; Institute of Electrical and Electroni: En;ineers (IEEE) Standards.

IEEE Std 279-1968 "Critaria for Nuclear Pouer Plant Prot 2ction Systers" IEEE Std 279-1971

" Criteria for Protection Systens for 1.2 clear Power Generating Stat ans."

IEEE Std 308-1971 "Crite:La for Class IE Electrical Systers fot Nuclear Power Generating Stations."

IEEE Std 317-1972

" Ele;trical Penetration Assenblics in Cont'ain:2ut Structures for Nuclear Fueled Power Generatin; Stations."

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d 323-1971 "IEEE Standard for Qualifying Class IE Equipment for Nuclear Pouer Generating Statiens."

!EEE Std 334-1971

" Trial-Use Guide for Type Tests of Centinacus-Duty Class I Motors Installed Inside tha Con

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iuclear Power Generating S tations. "

IEEE Std 336-1971

" Installation, Inspection and T2 sting Require.. ants fc? Instrumentation and Electric Equipment During tha Con-s cractioc of Nuclar Power Generating Stations."

IEEE Sti 33f-19.'1

" Trial-Use Criteria for the Periodic Testias o f iucl2ar Pouer Generating Station Pro tecti'te Sys tats."

IEEE s td 3 ^.-13 71

" Trial-Use Guida f or Seismic qualification o f Class I _actric Equipment for Nuclear Power Generating Stations."

IEEE ' d 130-1972 "IEEE Recontendad Practice for :laintenance, Tusting and Replacement of Large Stationary Type Power ?lant and Schstation Lead Storage Batteri2s."

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