ML20039G531

Forwards Draft Errata to Be Incorporated Into Next Suppl of Ser.Errata Addresses Section 9.5.5 Re Fire Protection for Specific Areas,Section 8.2 Re Offsite Power Sys & NUREG-0737 Item II.F.2 Re Detection of Inadequate Core Cooling

ML20039G531
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	01/13/1982
From:	James Smith LONG ISLAND LIGHTING CO.
To:	Joshua Wilson Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.D.1, TASK-2.F.2, TASK-TM SNRC-661, NUDOCS 8201180380
Download: ML20039G531 (29)

Text

{{#Wiki_filter:. x ^ LONG ISLAND LIGHTING COM PANY faafysw SHOREHAM NUCLEAR POWER STATIOBl** P.O. BOX 618, NORTH COUNTRY ROAD + WADING RIVER, N. .11792 January 13, 1982 SNRC-661 N A Mr. J. N. Wilson / Division of Licensing f }lbCIN N O g U.S. Nuclear Regulatory Commission L Jg"gI5f982m. Washington, DC 20555 d answ%augig ga Shoreham Nuclear Power Station - Unit 1 nu Docket No. 50-322 q e

Dear Mr. Wilson:

Enclosed are suggested errata which, it is hoped, can be incorporated into the next supplement to the SER. These errata are divided into three parts:

1) - Errata for the original SER

2) - Errata for SER Supplement No. 1

3) - Errata for Appendix C of SER Supplement No.

1. The input provided in Attachments 1 and 2 consists of marked-up pages of the SER and SSER No. 1 (except for Fig. 17-1), The input provided in Attachment 3 is in a different format and has, therefore, been segregated. Of particular concern is page 9-10 of Attachment 2, regard-ing clarification for the "20 ft. separation between redundant safety-related components". Clarification of this statement was first made in a telecon on 7/28/81, a copy of which is attached for your convenience. Should you have any questions, or require additional informa-tion, please do not hesitate to contact this office. Very truly yours, .0' J. L. Smith Manager, Special Projects Shoreham Nuclear Power Station RWG:mp d Enclosure cc: J. Higgins 8201180390 820113 FC'893 5 PDR ADOCK 05000322 o PDR

AHached i 8 ELECTRIC POWER 8.1 Gealeral The bases for our evaluation of the applicant's designs, design criteria, and design bases for the Shoreham electric power systems are set forth in the Standard Review Plan (SRP) (NUREG-75/OS7) Section 8.1, Table 8-1, " Acceptance Criteria for Electric Power." These acceptance criteria include the applicable General Design Criteria (Appendix A to 10 CFR Part 50), and reccmeendations and guidelines of regulatory guides, branch technical positions, and industry standards. We have determined that conformance to the applicable general design criteria and recc.n endations and guidelines cited above provides sufficient bases for acceptance of the electric power systems. It should be noted that we are currently reviewing the applicant's fire hazard analysis and the results of our fire review may have an impact on some aspects of the electric power system designs that we have reviewed and found acceptable. We will report on the results cf our fire review and the impact en cur previous design review findings in a supplement to this report (see Section 9.5). 8.2 Offsite Power System In order to satisfy the requirements of General Design Criterion 17, the design provides for an offsite electric power system and an onsite electric power system. The electrical grid is the source of energy for the offsite power system. The safety function of the offsite power system (assuling the onsite power system is not functioning) is to provide sufficient capacity and capability to assure that the structures, systems and components important to safety, perform as intended. The objectives of our review were to determine if the offsite power system satisfies the criteria set forth in Section 8.1 of this report and can reliably perform the functions that are assumed and used as a bases in the accident analyses for normal and abnormal plant conditions. Shoreham will be interconnected to the electrical grid system through four 138-kilovolt and one 69-kilovolt transmission lines terminating ct a 138-kilovolt switchyard and a 69-kilovolt transaission system. The two types of high voltage transmission lines converge on their respective locations in the plant through separate and independent routes. The 138-kilovolt switchyard is arranged in an "I" bus configuration (two buses joined at a common point), with the provi-sions to be converted to a combination of "I" - ring bus arrangement with the addition of another connecting bus. The 69 kilovolt transmission system includes an onsite 55-megawatt gas turbine generator. Power from the Shoreham main generator is supplied to the 138-kilevolt switchyard via a double circuit con-taining two half-size main stepup three phase transformers and two circuit breakers. TkG 13 Kv sJddva. J and -tto 44 Ye _s Asfahe.n emel has Ns owes The switchyard outerv svston consists nf two senarne1125-volt battery-bank charger un1Ls and attencant distribution systens. These pcwer sources provide cortrol cower to the breakers in the 133-kilovolt switchyard and 69-kilovolt (transm ssica systea} gv6gfay,*,,,, n$GfenYCn$ tourct's ef g-1 Sfr? w

m____.-.m--__- ~ _ - - I Il RADIO /,CTIVE WASTE MANAGEMENT 11.1 Su~. mary Description The radioactive waste management systems are designed to provide for the con-trolled handling and treatment of liquid, gaseous, and solid radioactive mate-rials generated during normal operation including anticipated operational occur-The liquid waste treatment system processes liquid from such sources rences. as equipment drains, systen leakace, condensate,.deminer lizer regenerant solu-tions, laboratory and decontamination liquids, and detergent wastes. The liquid waste is processed and recycled for reuse if the plant water balance requires makeup and the water quality is adequate or monitored and released to Long Island Sound. Gaseous wastes consist of offgases frca the main condenser air ejector, vents from equipment containing radioactive materials and leakage from systems and components containing radioactive material that are released through the building The offgases from the main condenser air ejector are treated ventiletion systems. by catalytic recombination to reduce the volume of of fgases and by charcoal adsorption to selectively delay fission product ncbie gases before release to the environment. Certain equipment vents and building ventilation exhausts are treated by high efficiency particulate air fiiters and charcoal adsorbers to remove radioactive particulates and radiciodine prior to release atmosphere. 4tr ceN6, Solid stes generated during plant operation consist of spent demineralizert resin evaporator bottoms, discarded radiocctive components and tos h, andj miscellaneous dry solid wastes. Wet solid wastes will be solidificc u e crummed. Dry, compressible materials will be compacted and drummed. Drummed wastes will be shipped to a licensed burial site. In our evaluation of the liquid and gaseous radicactive waste treatment systems, the following areas were considered: (1) the ca:a:,ility of the liquid and gaseous waste treatment systems to keep the levels of ra:icactive material in effluents "as low as is reasonably achievable" based on ex:t:ted radwaste generated over the life of the plant, (2) the capability of the 'iquid and gaseous waste treatment systems to maintain releases below the.imits specified in 10 CFR Part 20 assuming fission product leakage at desir levels from the fuel, (3) the capability of the liquid, gaseous, and solic -aste systems to meet the processing demands of the plant during anticipata: cperational occurrences, (4) the potential for gaseous release due to hyc gen explosions in the gaseous radwaste system, (5) the design features incorp;rited to control the release of liquids c,ntaining radioactive materials due n tank overflows, and (6) the provisions for monitoring and controlling radicinive materials in process and effluent streams and the provisions to detect lair. age of radioactive material between systems. In our avaluation of the radioactive waste mana nent system, we considered the inforraation provided in the Shoreham Final lif ety Analysis Report and the document, " Compliance with 10 CFR 50 Appendix I,' submitted by Long Island 11-1 cri'

Table 11-4 Design Parameters of Principal Components Considered in Radwaste Evaluation Liquid Systems Number of Capacity Quality Seismic Components Each Component (each) Group Design Waste Collector Tank 2 25,000 gallons a b Radwaste Filter 2 flO / b square feet a b (150 gpm) Recovery Sample Tank 2 25,000 gallons a b Radwaste Demineralizer 2 48 cubic feet a b (150 gpm) Discharge Waste Sample Tank 2 25,000 gallons a b Floor Drain Collector Tank 2 25,000 gallons a b Waste Evaporator 1 20 gallons / min. a b Regenerant Liquid & EvaporatorTank 2 25,000 gallons a b 2A/5[2M00)ga11ons ~ b a Evaporator Bottoms Tank 1 Cleanup Phase Separator 2 4,700 gallons a b Spent Resin Tank 1 4,700 gallons a b Regenerant Evaporator 1 20 gallons / min. a b Laundry Drain Tank 2 1,600 gallons a b fff(77squarefeet a b Laundry Filter 1 (50 gpm) M quare feet a b Floor Drain Filter 1 33,# (50 gpm) O t 11-7 _ m ___ m,w _. m,

Influents to this subsystem include liquid wastes from the spent resin storage tank and low conductivity wastes from the condensate demineralizer wash water. These are cc,llected in one of two 25,000 gallon waste collector tanks, processed in a batch through one of the two 150 gallon per minute precoat filters and radwaste demineralizers, and collected in one of two 25,000 gallon recovery From the recovery tanks, the liquid may be sent to the condensate storage tanks. tank or routed ta the discharge waste sample tanks for release at a controlled rate to the discharge canal. Wastes to the condensate demineralizer regenerant chemical subsystem are collected The in one of two 25,000 gallon regenerant liquid and evaporator feed tanks. I collected wastes are neutralized, and processed through a 20 gallon per minute j Condensate from the evaporator is sampled and recycled regenerant evaporator. to the waste collector subsystem for use in the plant or routed to one of two n IT ' 25,000 gallon discharge waste sample tanks, depending on the plant water balance r )h and the quality of the condensate. The floor drain (high conductivity) subsystem collects liquids from the drywell, G containment, reactor building, fuel handling area, radwaste building and turbine ,y building floor sumos, laboratory drains and nondetergent decontamination solutions. O Ic i( The wastes _are.collec_ted in one of two 25,000 gallon floor drain collector tanks, The filtetedland_ processed through a 20 gallon per minute waste evaporator. [evaporatorcondensateissampledandrecycledtothewastecollectorsubsystem If the l0 N for use in the plant or routed to the discharge waste sample tanks. conden .( j%E; discharge canal at a controlled rate from the discharge waste tanks. Laboratory wash water and wastes containing detergents (laundry, personnel and Tj equipment decontamination wastes) is collected in one of two 1600 gallon laundry These wastes are filtered and routed through the floor drain filter gi drain tanks. gl' to the discharge waste sample tanks or treated by the waste evaporator. 4-b. The applicant estimated annual average input rates for the waste co I C' n The / per day, 1550 gallons per day, 10,000 gallons per day, respectively. applicant considered that approximately 10% of these treated wastes would be [V) ' \\. For the detergent wastes, the applicant estimated 500 gallons per discharged. day and considered 100% of these wastes would be discharged. In our analysis, S based on the parameters given in NUREG-0016, we determined the rates for the Q waste collector, regenerant chemical and floor drain subsystems to be 29,300 ( t gallons per day, 1700 gallons per day and 6300 gallons per day, respectively, c We assumed A with 90% of the treated wastes recycled for reuse in the plant. that450gallonsperdayofdetergentwastesareproduced, filtered,andreleasedQ. to the discharge canal. The design capacity for processing the waste collector N is 210,000 gallons per day. The design capacity for processing both_the regener ; ant chemical and the floor drain (plus detergent waste) subsystems (are 28,800 ' } a- _ gallons _per dayj The difference between the expected flows and the design flows. The waste collector will provide reserve capacity for processing surge flows. ~ subsystem demineralizers may be operated in a series or parallel flow, through normally.only one demineralizer will be on line. \\ Bottoms from the waste evaporator and the regenerant evaporator will be collected i in a 2500 gallon evaporator bottoms tank and transferred to the solid _ waste ,lclc c.S flo s/ ch 00 A bain s bf+N!raf.cn u?')' s go y oc w ~ &-9 w %. ,/ 11 'N_ m _~ f 2 M R *' Y _ m. my y y

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7 t bcle h c.Jalp. Mo son ry cement w;ll Le us e d. Wet solid wastes will consist mainly of spent demineralizer resins collected in 4700 gallon spent resin storage tank and waste evaporator bottoms from th_e q Wetsolidwastewillbemixedwithfcatal t/ "/and> solidifying agent (cement) and transferred to 100 cubic foot tran 7@0 gallon waste dewatering tank. Prior shipping containers and shipped to an offsite licensed burial site. testing will be performed to determine the waste solidification ratios in the cement waste mixer to assure proper solidification. The applicant has committed to sample each liner of air-dried filter cake discharged from the floor drain and radwaste filters prior to packaging in transport shipping containers. If5 the contents do not qualify as low specific activity material and is 6 x 10 millicuries per gram, or greater, the raterial will be solidified in accordance with Branch Technical Position ETSB 11-3 using a portable solidification system in the radwaste building. We find the applicant's proposed method for packaging filter cake to be acceptable. Dry solid wastes will consist of ventilation air filters, contaminated clothing, paper, and miscellaneous items, such as tools and laboratory glassware. Dry solid wastes will be compressed into 55 gallon drums by using a baling machine. The baling machine is equipped with a ventilation shroud that controls the release of airborne dust by an efficiency filter on the exhaust. Based on data on solid wastes from operating BWRs, the staff estimates that approximately 31,000 cubic feet of wet solid wastes containing approximately 3500 curies and appoximately 14,000 cubic feet of dry solid wastes containing less than five curies will be shipped off-site annually to a licensed burial site from Shoreham Nuclear Power Station. The principal isotopes expected to be present in these wastes are Cs-134, Cs-137, Cr-51, Co-58, Co-60, Fe-55 and Mn-54. The applicant's radwaste building design provides approximately thirty days onsite storage of packaged wastes prior to shipping. Our estimate considers the applicant's design to provide 1500 square feet storage area at ground level near the loading dock in the radwaste building. Our review included an evaluation of the system's capability for processing types and volumes of wastes expected during normal operation including anticipated operational occurrences. The management, process control and design of the solid radwaste system equip-ment and instrumentation meets the guidelines of Branch Technical Position ETSB 11-3, " Design Guidance for Solid Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants." The solid radwaste ~ treatment system will be located in the radwaste building, a seismic Category I structure. The seismic design and quality group classifications of the equipment meets the guidelines given in Regulatory Guide 1.143. All containers are required to be shipped to licensed burial sites in accordance with Nuclear Regulatory Commission and Department of Transportation regulations. 11.5 Process and Effluent Radiological Monitoring Systems The process and effluent radiological monitoring systems are designed to provide information to operations personnel on radiation levels in the plant process I i 11-13 l &M = . Mabi4 o lW% a h .a w W= m% J 9' am.*.*e.a4 h h 4.-I4,- 4s-a4 h a eb ..a s A4 em h-e=e

Th t:. fan!e cl odd 56 ey,3n deol l0 inc lscl2) g// arm Q /,j,./ on FS A I'i' B bleS H. +. 2 - / A, 8 an d // 4. 3 - /, - Table 11-5 Process and Effluent Radiological Monitoring System

• Number Monitor Monitor Stream Monitored and Type Style Sensitivity Liquid *

-6 Liquid Waste Effluent 1-offline y Scintillator 10 p Ci/ml (Cs-137) -6 Residual Heat Removal 2-offline y Scintillator 10 p Ci/ml Service Water (Cs-137) -6 Reactor Bldg. Closed 1-offline y Scinti11ator 10 p C1/mi Loop Cooling (Cs-137) -6 Radwaste Bldg. Closed 1-offline y Scintillator 10 p Ci/ml Loop Cooling Outlet (Cs-137) -6 Radwaste Stea:n 1-inline y Scintillator 10 Ci/mi Generator Outlet (N-16) -6 Steam Seal Evaporator 1-inline y Scintillator 10 p Ci/ml Outlet (N-16) Gaseous -6 Station Ventilation 1-offline p Scintillator 10 p Ci/ml Exhaust (Gas) (Kr-85) Station Ventilation -6 Exhaust (Particulate)*** 1-offline p Scintillator 10 p Ci/ml (I-131) -6 Radwaste Bldg. 1-offline p Scintillator 10 p Ci/ml Tank Vent (Kr-85) Offgas Vent ** 1-offline y Scintillator 1 mr/hr Discharge -6 Condenser Vacuum 1-offline E Scintillator 10 p Ci/mi Pump Discharge (Kr-85) -6 Standby Ventilation 2-offline E Scintillator 10 p Ci/ml Exhaust (Kr-85) All liquid and gaseous streams will be monitored in accordance with the guidelines of Regulatory Guide 1.21.

• • • Particulate filter and iodine samples are continuously collected and periodically analyzed.

11-15 w-TT " EE _,c .__.2 1

. b' go 17 QUALITY ASSURANCE f 17.1 General The description of the quality assurance program for the operations phase of the Shoreham Nuclear Power Station is contained in Section 17.2 of the Final Our evalation of this quality assurance program is Safety Analysis Report. based on review of this inforcation and discussions with representatives of the Long Island Lighting Company. We assessed Long Island Lighting Company's quality assurance program for the operations phase to determine whether it complies with the requirements of Appendix B to 10 CFR Part 50, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants"; the applicable quality-assurance-related Regulatory Guides and ANSI Standards listed. in Table 17-1; and the Standard Review Plan, Section 17.2, Rev. O, dated November 24, 1975, " Quality Assurance Ouring the Operations Phase." 17.2 Organization for the Ouality Assurance Proaram The structure of the organization responsible for the operation of Shoreham and for the establishment and execution of the enerations phase cuality assurance . program is shown in Ficure 17-1. l The Vice President of Operations has been delegated overall respons1Dility for the safe and reliable operation of all Long Island Lighting Company generating plants. He is responsible for the establishment, implementation, and effectiveness of the operational quality assurance program, policies, goals, and objectives. These responsibilities are carried out through: the Quality Assurance llanager, Manager - Electric Production-Nuclear, the Plant Manager, and the Director of Production. V Thehalityh;suranceDepartmentissupervisedbytheQualityAssuranceManager. "; report; functionalis ar.J ad.cinistcstivcly to the Scnicr Vice Prc ident, @o- -Engincciing and T.; W henges a;. ducin iii cnd fanc- -tienelly te the V, ice Pree dent of Cpercc'g r.sdificatica act v t c: icn3 fcr dicccticn during cp;rcting kS.cd pheac cctivitica.i[e. hc plity ?.;;urrm; 'b=pr is responsible for establishing * 's # impl:m:nting the Long Island Lighting Company quality assurance program at anoreham as described in the Long Island Lighting Company Quality Assurance Hanual. He is assisted in carrying out his responsibilities by onsite and offsite quality assur nce personnel. CLmM/ndp (CGi.Aw of t The quality assurance organization has the authority to: (1) identify quality problems, (2) initiate, recommend, or provide solutions, (3) verify implementation of solutions, and (4) stop unsatisfactory work or further processing of unsatis-factory material. The quality assurance organization is responsible for: (1) reviewing and approving quality related documents (e.g., instructions, procedures, and specifications), (2) performing vendor quality assurance prequalifications, (3) assuring that procurement documents contain quality requirements which can be inspected and controlled, (4) survc!11ance and auditing of vendors, (5) documenting and reporting to management noncamformances discovered during surveillance or audit, (6) assuring corrective actions are effective and acccmplishad in a timely manner, and (7) auditing of maintenance and opera-tion activities. 17-1

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e O INSERT A The Corporate Statement of Quality Assurance Policy assigns overall responsibility for the Quality Assurance Program to the Vice President, Engineering. The Vice President, Engineering is responsible for the establishment of QA policies, goals and objectires during operation of the nuclear power plant, and for assuring the implementation and effectiveness of the QA Program. He has delegated specific QA. Program responsi-bilitien to the Manager, Quality Assurance Department. The Vice President, Nuclear, has everall responsibility for the safe and reliable operation of the nuclear power plant. He is responsible for assuring the implementation of the QA Program requirements by the organizations under his jurisdiction. He has delegated to the Plant Manager the responsibility to assure inple-mentation of the LILCO QA requirements within the station.

l L 9psinut The Plant Manager reports to thc/,." = gar El=tric "rd= tierNuclear. He is i directly responsible for the safe and reliable operation of Shoreham and for assuring the implementation of the quality assurance program on site. I The Operating Quality Assurance Engineer reports directly to the Plant Manager t and maintains a working interface and a direct line of communication with the l Quality Assurance Manager. The Operating Quality Assurance Engineer has direct i respensibility for assuring implementation of the Long Island Lghting Company l quality assurance program and additions and changes thereto onsite. He is responsible for establishing, reviewing, and implementing all site quality assurance / quality control procedures and instructions and for the performance of site audits. He independently evaluates and reports the status and adequacy i 7f the quality assurance program at the station to the Plant Manager and the Quality Assurance Manager. Differences of opinion between the Operating Quality Assurance Engineer and Plant Manager involving quality matters will be referred to tne Quality Assurance Manager, and it necessary, to the highest level of manacement for ultimate resolution. 17.3 Quality Assurance Program i 1 The cuality assurance program description for the operation of Shoreham is k descriced in the Long Island Lighting Company Quality Assurance Manual and is supplemented by quality assurance procedures and instructions which provide the cetailed instructions and checklists necessary to impicment or verify } imple entation of the quality assurance program requirements. 1 The c.:ality assurance prograr for Long Island Lighting Company is structured i to be in accordance with Appendix B to 10 CFR Part 50 and with the provisions of t7e NRC regulatory guidance shown in Table 17-1. These documents, coupled with the quality assurance program description in the Final Safety Analysis Repe%, form the foundation from which the overall quality assurance program is fem.ulated and describe how the requirements of Aopendix B to 10 CFR 50 are satisfied. The program is implecented via the Long Island Lighting Company Quali ty Assurance Manual and implementing procedures. The Manual is approved by t'e Senior Vice President - T & D and Operations and the Senior Vice Presi:ent - Engineering and4Prf=t '!=v=t and is supplemented by quality 4 assura ce procedures and ins (tructions. These documents control quality-i rela:sd activities involving' safety-related items to satisfy the requirements of Aprendix B to 10 CFR 50.,g g g The c'fsite sperati= 1 quality assurance procedures and instructions are appre.ed by the Quality Assurance Manager. The ;tation quality assurance / quali:y control procedures and instructions are reviewed by the Operating Quali:p Assurance Engineer and the Quality AssuranceCr.;N.:r, approved by the Plan: Manager and issued by the Operating Quality Assurance Engineer. The quality assurance program requires that quality assurance documents encompass detai'ed controls for: (1) translating codes, standards, and regulatory require-ments into specifications, procedures and instructions, (2) developing, reviewin9 and a:croving procurement documents, including changes, (3) prescribing all quality-af fecting activities by documented instructions, procedures, or drawings. (4) i ssuing and distributing approved documents, (5) purchasing items and servi:es, (6) identifying materials, parts, and components, (7) performing j specid processes, (8) inspecting and/or testing caterial, equipment, processes I f 17-4 sre a _~

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Y ac menf 2 g 6.2.1.6 Subccrpartment Pressure Analysis l In the Safety Evaluation Report, we stated that the applicant would be required ~ to provide auditional information to allow us to complete our review of the The appli-forces on the reactor pressure vessel affecting the support skirt. since provided es with the results of an analysis that used the peak cant har pressure calculated by the staff in the determination of the moment on the reactor vessel skirt. This analysis resulted in an increase of 1.5% in the maximum asymmetric load on the support skirt. We find the rethods and assump-tions used by the applicant to be acceptable for calculating the pressure response between the reactor vessel and sacrificial shield in evaluating the design of the support skirt. 6.2.1.7 Steam Bypass of the Suppression Pool _ In the Safety Evaluation Report, we stated that the acceptance criterion for the periodic (every refueling outage) low pressure tests must be 20% of[f4e jdan7' The applicant has subsequently submitted a 2 bypass __capabilitypf;0.05ft. canalysis of the steam bypass capability using new methods of accounting for the mass and energy transfer between the suppression pool and the containment These methods have not been previously used for licensing purposes atmosphere. and will require a considerable amount of time for the staff to complete its '+ h]~!Yi&~N R ??]: '

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review. c> Since J e issue here is an_ acceptable criterion for.(a pressure test, t n{t'be perform;ed yntiTthe first refFeTing outage; it is the staff's position e I I t at~ 3 e resolution of this topic be a License Condition, to be completed g { before the start of the first low pressure test of the drywell. i The applicant has committed to use 10% of the allowable by?3ss capability as This meets the the acceptance criterica for the periodic low pressure tests. staff's position, and is therefore acceptable.

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'I The applicant has stated that the vacuum breakers will be leak tested by one of two possible methods and that this leakage will be added to the overall by-The pass leakage measured in the single preoperational high pressure test. two methods under consideration are to expose the downstream side of the vacuum breakers to the test pressure or to test them separately. Both methods are acceptable to the staff. 4 6.2.1.8 Pool Dynamics In the Safety Evaluation Report, we stated that our review of some of the Mark II pool dynamic loads was continuing. This supplement presents our evaluation l' of the proposed load specifications for Shoreham that were identified as l l-outstanding items relative to the Generic Acceptance Cr;teria developed within Task A-8, " Mark II Containment Pool Dynamic Loads," and Task A-39, "Determina-j,. tion of Safety Relief Valves (SRV) Pool Dynamic Loads and Temperature Limits l - > J for BWR Containments." ll i ,J t IiY p ^N -u-- c. m ma% _ s

1. Relay Room 2. Control Room 3. Computer Room 4. Emergency Switchgear Rooms 5. Battery Rooms In ducts penetrating the fire barrier walls surrounding the safety-related equipment, a fire damper of 1\\-hour rating is used. Some areas also contain motorized 1 -hour fire dampers in which tne motorized assembly, including cables, are not U.L. listed. We are concerned that the unlisted assemblies will prevent the fire dampers from performing its function. We require that all such operators be replaced with approved listed operators or a surveillance program be developed and included in the plant Technical Specifications to assure an adequate level of reliability. 9.5.4 Emeraency Lighting Eight-hour battery pack emergency lights are required for areas of the plant necessary for safe shutdown. The applicant will install self-contained eight-hour battery pack emergency lighting in all areas of the plant which could be manned to bring the plant to a safe cold shutdown and in access and egress routes to and from sl Ffire' areas. -c We conclude that tie emergency lighting meets the requirements of Appendix to BTP ASB 9.5-1, and, also, the previsions of Section III.J of. Appendix R to 10 CFR Part 50 and is, therefore, acceptaale. l 9.5.5 Fire Protection for Soecific Areas 9.5.5.1 Control Room The control room complex is separated from all other areas of the plant by 3-hour fire rated walls, ceiling / floors assemblies, floors and doors. All ventilation ducts penetrating these barriers have 3-hour fire rated dampers. The control room complex peripheral rects, except the visitors gallery which l has bullet-resistant noncombustible materials, are constructed to provide a minimum fire rating of 1 hour. The ventilation openings in the peripheral rooms are protected with 1 -hour rated fire dampers. At our request the applicant has agreed to install additicnal smoke detectors in these rooms which will alarm and annunciate in the control room. All cabinets, consoles, and the ventilation exhaust system within the control l ' room have ionization fire detectors installed. The main control room ventila-tion system can be remote manual isolated from the main control room as it has capability of being used as a smoke removal system. Manual fire fighting is provided through the use of portable extinguishers and CO hose reels (supplied from the stati:n hose pressure C0f storage tank) 2 which are located outside the main centrol panel at the access door. At our 9-7 4 m.a .:wo.. m._. rm,,_ _

.~ separation analysis of cables witt.in the reactor building by letter dated February 10, 1981 and analysis of shutdown circuits outside the reactor building by letter dated July 10, 1981. The applicant's post-fire safe shutdown ar.alysis demonstrated that systems needed for hot shutdown and cold shutdown are redundant and that one of the redundant systems needed for safe shutdown would be free of fire damage, by providing separation, fire barriers, and/or alternative shutdown capability. The safe shutdown analysis considered components, cabling, and support equipment. for systems needed to shut down. Thus, in the event of a fire, at least one train of systems free of fire damage would be available to achieve and main-tain het shutdown or to proceed to cola shutdown. For hot shutdown, at least one of the follcwing shutdown systems would be available: (1) the Reactor Core Isolation Cooling System, (2) the High Pressure Coolant Injection System, and (3) a combination of the pressure relief system, the core spray system and residual heat removal (RHR) system. For cold shutdown, an appropriate portion of the RHR system would be available. For equipment located in the primary containment, no fire protection features are provided because the containment atmosphere will be inert. For equipment located in the reactor building (secondary containment), the applicant provided a cable separation analysis which divided the reactor building into overlapping 45 cegree segments. The applicant assumed that all components, the cables and raceways, in a given segment were lost due to a fire; yet demonstrated the capability to shut down still existed. We have reviewed the cable separation analysis and conclude that it is an acceptable method of demonstrating that adaquate separation exists between the redundant trains. Additionally, the applicant has committed (by July 10, 1981 letter) has a minimum 20._f taepatation_betweert to verify that the "as-built" design,G d e ca h r o redundant sale.ty rJ_,l_ ale _dEmponentsx 66w do cuhM J Q y ys o c y

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The secondary containrent is a cylindrical structure with a 135-foot outside diameter and 240 ft high with 2-foot concrete walls. There are six complete l elevations with each elevation containing large open penetrations. The area i between the primary and secondary containment is one fire area.

wHd ver+,bal Throughout the reactor (building both smoke and temperature detector? are l

installed with alarm anchannunciation in the control room._ AllEible trays [n the have solid bottoms withAcovers.[oFladder typ~e With sol'id covers attached ~ ~~ ~~ e ca c-rce j@ external to all trays where they penetrate floor levelsh_ sides 3 All vertic ~ ~~~ ~ Fire stops are a provided at the midpoints when the elevation is more than 25 ft. I The two main vertical safety-related cable risers are located at 138 and 223 azimuth, extend 9am elevation 8' to elevation 40' and are separated by 85 ft. i The applicant's analysis demonstrated that a 45 segment in which a fire I caused the disability of all cabies and raceways in that segment, a separation i distance of 20 ft on the inside of secondary containment existed and 35 ft I existed on the outside. The applicant then rotated this segment 22.5 for additional verification and overlapping. 9 9-10 3 e r-

The applicant provided fire detection, alarm, annunciation, water spray systems for the RBCV's charcoal filters, hose stations, automatic pre-action sprinkler I systems, fire barriers, two fire main feeds and portable ecuipment for secondary containment. Due to the preceding separation distances and protection provided, an automatic suppression system is not needed for protection against a transient exposure fire. For equipment in areas outside the reactor building, the applicant has identified seven areas which contain cable for redundant shutdown equipment: the relay room, the control room, the diesel generator rooms, the emergency switchgear room, the fuel oil pumphouse rooms, the screenwell, and the HVAC roon. In the diesel generator rooms, the emergency switchgear ro'om, the fuel oil pumphouse rooms, and the screenwell, recuncant equibment is separated by a 3-hour fire rated barrier. Cabling to this equipment is contained in underground ducts. In the event that fire disables redundant equipment in the HVAC room, control room, or relay room, a remote shutdown panel is provided in the reactor building (refer to section 7.4.3 of this report). l (ccjac,jlYy p Sections 7.4.1.4, 7.5.1.4, and 7.5.1.5 of the FSAR describe the' remote shutdown panel's design and capability. By letter dated May ll, 1951,fthe applicant addressed Section III.L of Appendix R. The design obje'ctiTe'of the remote -~ shutdown panel is to achieve and maintain cold shutif5snTin e~v'en~t' o'f a fire disabling the relay room or the control rccm. The reactoTr6re ~is'olation cooling (RCIC) system, safety / relief valves and one divisionfof the residual heat removal (RHR) system can be controlled from the remote shutdown panel to.x s achieve cold shutdown. E b y, p e s.c Q g o n t. geilj ad irrtM'o f~o ta cj The design of the remote shutdow(n'paTrel comphes witn tlie performance goa wob r n-- n a n ey 3 outlined in Section III.L. Reactivity control will be accomplished by a manual scram before the operator leaves the control room. The RCIC system will provide reactor coolant makeup and the RHR system and the safety relief valves will be used for reactor heat removal. Reactor water level, reactor pressure, suppression pool water level and temperature ~, and drywell pressure and temperature are among instrumentation available at the remote shutdown panel to provide direct reading of process variables. The remote shutdown panel will also include instrumentation and control of support functions needed for the shutdown equipment. Procedures for use of the remote shutdown panel include sequencing of equipment and operator actions. Based on the above, we conclude that the fire protection of safe shutdown capability meets the guidelines of Appendix A to BTP 9.5-1 and is, therefore, acceptable. 9.5.7 Administrative Controls and Fire Bricade The administrative controls for fire protection consist of the fire protection organization, the fire brigade training, the controls over combustibles and ignition source, the prefire plans and procedures for fighting fires and quality assurance. The fire brigade will be composed of five members per shift. To have proper coverage during all phases of operation, members of each shift crew will be trained in fire protection in accordance with our guidance including Regulatory Guice 1.101, " Emergency Planning for Nuclear 9-11 G,.__ m. .__. e m w.m e L A

i I 1 I of 10 CFR Part 50 to base its finding on a review of FEMA's findings and deter-minations as to whether State and local emergency plans are adequate and capable of being implemented, and on the NRC assessment as to whether the applicant's onsite emergency plans are adequate and capabic of being imple- -mented, the following items are required in order to make a favorable finding. i I A. Correction of the deficiencies identified in the foregoing evaluation to an extent sufficient that the 16 standards in Section 50.47(b) of 10 CFR Part 50 are met. B. Submission of the State and local emergency response plans in accordance with Section 50.33 of 10 CFR Part 50. The submission should include a definitive description of the Emergency Planning Zones (EPZs) including the factors identified in the aforementioned section of the regulations which were used in determining the exact size and configuration of the EPZs. C. Submission and satisfactory evaluation of the emergency plan implementing procedures in accordance with the requirements of Section V of Appendix E to 10 C 7 Part 50. i D.' Acceptable review of the FEMA findings and determinations as to whether the State and local emergency response plans are adequate. E. Acceptable findings from an onsite appraisel to establish that the appli-cant's plan is capable of being implemented. F. Acceptable completion of the joint exercise held in accordance with the 1 requirements of Section IV.F.1.b of Appendix E to 10 CFR Part 50 to j demonstrate the dynamic capability to implement the onsite and offsite emergency response plans. Upon completion of the above items, the staff's evaluation of the overall state of emergency preparedness for the Shoreham site will be presented in a supplement to the Safety Evaluation Report. 13.4 Review and Audit The applicant has made and is making provisions for the review and audit of j. plant safety-related activities, as detailed below. 13.4.1 Review of Ooerations Committee (ROC) The Review of Operations Committee has been establ1<hed and is-functioning during the completion of plant construction. The 4CO as currently constituted is described below, gg,efe,, In enter 80C The Plant Manager is the Chairman of the ROC. The other members are the Chief Operating Engineer (Vice Chairman), Chief Technical Engineer, Operations Engineer, Maintenance Engineer, Ins _rument and Control Engineer, Radicchemistry Engineer, Health Physics Enginee; and Operating Quality Assurance Engineer. A quorum consists of at least five members or their designated alternate and includes the Chairman or a Vice Chairman. Meetings of the ROC are held monthly. 13-38 w

close proximity to recirculating fluid systems inside and outside of containment. Non-LOCA events both inside and outside of containment should use 10 percent noble gases, 10 percent iodines, and 0 percent particulate as a source term. The following table summarizes these considerations: LOCA Source Term Non-LOCA (Noble Gas / Iodine / High-Energy Line Break Source Term Containment Particulate) (Noble Gas / Iodine / Particulate) Outside Percent Percent (100/50/1) (10/10/0) l in reactor in reactor coolant system coolant system Inside Larger of (10/10/0) (100/50/1) In reactor in containment coolant system Ef (100/50/1) in reactor coolant system Discussion and Conc.1,usions The applicant utilizeo uur prescribed post-accident source terms described in Regulatory Guide 1.3 and 1.7 (as specified in NUREG-0737) to perform its radiation and shielding design review. The applicant considered the following two accident scenarios, a LOCA, and a pipe break in the secondary containment, along with the above mentioned source terms, to calculate the in-plant post-accident radiation levels at Shoreham. The radiation sources considered in calculating these dose rates included; 1) direct radiation from airborne and liquidborne radioactivity in the primary (drywell) and secondary containment (Reactor Building); 2) direct radiation shine from system piping and components in the primary and secondary containment; and 3) immersion dose rates from air-borne radioactivity due to primary containment (and equ'ipm@dakage. As specified in NUREG-0737, the applicant has identified the plant systems which will be required to function to mitigate a LOCA, and which may contain high levels of radioactivity. In performing the shielding review, the computer codes QAD-MOD and ANISN "D" were used to determine the post-accident dose rates at Shoreham. The control room, the Technical Support Center (TSC), and the Post Accident Sampling and Analyeis Facility (PASF) are all areas which will require con-tindous or frecuent occupancy following an accident at Shoreham and are considered vital areas. LILCO evaluated the other areas suggested as vital post-accident areas in NUREG-0737, but concluded that these areas either do not apply to Shoreham or their functions can be controlled / monitored from the main control rocm at Shoreham. The control room and permanent TSC will have post-accident dose rates of less than'15 mrem /hr (30-day average). The inte-grated dose to these areas for the duration of the accident will be less than 22-35 " tr ~_i m..,.

II.F.2 Instru entation for Detection of Inadequate Core Coolinq Position Licensees shall provide a description of any additional instrumentationo[ controls. (primary or backap) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide an unambigueus, casy-to-interpret indication of inadequate core coolir.g S (ICC). A description of the functional design requirements for the system shall also be incit:ded. A cascription of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided. Clarification '7-(1) Design of new instrumentation should orovide an unambiguous indication of ICC. This may require new measurements or a synthesis of existing measure-ments which meet design criteria (item 7). (2) The evaluation is to include reactor-water-level indication. (3) Licensees and applicants are required to provide the necessary design analysis to support the proposed finsi instrumentation system for inadequate core cooling and to evaluate the cerits of various instruments to monitor water level and to monitor other parameters indicative of core-cooling conditions. (4) The indication of ICC must be unambiguous in that it should have the [ 2 following properties: (a) It must indicate the existence of inadequate core cooling caused by 4 varicus phenomena fi.e., high-void fraction pumped flow as well as stagnant boil-off); and (b) It must not erroneously indicate ICC because of the preserce of an unrelated phenomenon. i (5) The indication must give advanced warning of the approach of ICC. I (6) The indication cust cover the full range from normal operation to complete core uncovery. For example, wc*.cr-level instrumentation may be chosen to l ~ provide advanced warning of two phase level drcp to the top of the core. l and could be supplemented by other indicators such as incore and core-exit thermocouples provided that the indicated temperatures can be correlated z to provide indication of the existence of ICC and to infer the extent of x 1 core uncovery. Alternatively, full-range level instrunentation to the bottom of the cara cay be employed in conjunction with other diverse indi-l cators such as core-exit thereccouples to preclude misinterpretation due to any inherent deficiencies or inaccuracies in the measurement system selected. g) =7j[gf (7) All instrumentation in the f'"C ICC system must be evaluated for conform-ance to Appendix B of NUREG-TU " Clarification of TMI Action Plan l Requirements," as clarifiec or nodified by the provisions of items 8 and 9 l that follow. This is a new requirement. 22-68 -me ..m_.,-~.,,:%- ..ev_,,,, ---.~._,-,_-,.,..-m,.r._m,. ,m.,

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' Tengled and kinked sound powered phon'c cords lying on the floor are a tripping

'- hazard. Use of non-kink or self retracting cords, is recon =anded. (1) t 1.1LCO Respcnse" s During'nor::pl cperation, 'the use of sound powered phones in the control roon is aocaliced at specific panals. Phone cords will be of appropriate icogth and pose l h,%, beneficiil and, in come cases, may even creat'e a hacardous condition. no tEippin'g hazard. LILCO does not believe a change 'n. cord type will be 4 4 -s. b

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......-,s..- [ 2.3 The cmargency DC lig4tir.g'in th'e bdck panel areas is inaderuste. Measured lighting illumigance ranged fro:2 less than one (1)ifcot-car.dle to seven '(7) foot-<andles.'N(2.4). ( s% t, LIlCO_Respan u. There pre f.htte independent lighting systems in the Shorcham Control.Rocm as described,below: (a) Normal AC. All fluorescent lighting fixtures, ee:tcept those directly above the inner ring panels, are pcwered by nornal AC, which is backed up by diesel generator 101 upon lo,5 of offsite pcwer. (b) Inverter AC.' The fluorescent lighting fixtures dire-tly over the inner ring panels are powered from the vital bus that will automati-cally transfer to diesel generator 103 upon loss of offsite pcwer. s (c) Emergency DC. There are approximately 12 incandescent light fixtures throughout the control rocm wnich are powered directly from the station batte,y. Based on the above, upon loss of offsite power, the fluorescent lignts directly \\ w"over the inner ring paneli will stay illuminated while the balance of fluorescent - fixtures will temporarily be 'off for about ten seconds until diesel generator 101 comes on line, at which time they will be re-energized. If a LOCA and a loss of offsite power occur concurrently, the light feeder breaker will be tripped and the operator _will have to manually override the trip function to reclose the treaker. + = In addition, S-hour battery pcwered lighting will be provided in the control ) rocm and access and egress routes prior to loading fuel. s. 3 b

WW Mt _ b b b~ h.. -._.. Wrk 0%5 .i NRC Findinp, 2.4 hN (( Q{ The escrgency DC lighting in the back panel areas is inadequate. Icasured . lighting illuminance ranged f rom less than one (1) foot. candle to seven (7) foot candles. (1) Recorcended: 30 foot candles miniana for safety eclated pancis and 7.5 foot candles cinimum for non-safety related panels. b _ LILCO Response l There are three independent lighting systccs in the Shorchan Control Room as described below: a) Nornal AC. All fluorescent lighting fixtures, execpt.those directly above the inner ring panels, are powered by nor:-al AC, which is backed ep by dicsc1 generator 101 upon lon.s of of fsite power. b) Inverter AC. The fluorescent lighting fixturas directly over the inner -ring panels are powered from the vital bus (inverter) which is powered froa cither station battery C, or an AC bus that will automatically transfer to diesel generator 103 upon loss of offsite power. c) Energency DC. There are approxicately 12 incondescent light fixtures throughou t the control room which are powered directly from the station bat te ry. Based on the above,- upon loss of of f cite power, the fluorescent lights-directly over '.he inner ring panels will stay illuninated uhile the balance of fluorescent 4 . h fixtures will temporarily be off for about ten seconds until diesel generator 101 l co cs ca line, at uhich time they will be re energiced. If a LOCA and loss of offuite power occur concurrently, the lighting feeder breaker uill be tripped and-the operator will have to c.:nually override the trip function to reclose the ( breake r. In addition, ei ht-hour battery pcwered lighting will be provided in the control d room and access and egress routes prior to fuel load. This lighting has been ~ g addressed in our responce to Appendix R,.100FR50. [ g, .e .g an-ab -w mim m m e-- h e m p e ami. ew -g y w A A a = M wd4A J l SS5k. l l

i f1RC Finding 5.13 S Q Q_, 58 5 There are cany instances of adjacent meters having different scales. (2) LILCO Response The scales on eeters in the control rocu are deternined by the variabic being monitored. The location of the meters is nornally determined by the systen with which they are associated. LIhCO believes this to b2 the most efficient method of operation. D. O t .m 6.12 Containment Purge Contro1 Valves 39 C and D are reversed with rc5cect to their depiction on the,imic scove. (6.23) 1ILCO Rescansa .L The valves in c,uestion (IT46*A0V33C, 380, 39C, and 390) are 18" suppres-sion chamber purgs valves. Prior to the decision to inert the Shoreham I containment, these valves were intended to purge the suppression chamber l periodically for hydrogen and pressure control. l With the feplementation of the containment inerting system and the addi-l tion of a 6" vent line (IT46*AOV73A and B), the above valves will not be g used during normal operation. Valves 38C, 380, 39C, and 39D will be used to purge the suppression enamber prior to entry during shutJown condition. Also, both valves have to be operated to perform the function and, from 3 an operating point of view, relative switch vs. nimic location has little meaning. i Each valve is individually key locked closed and under control of wera-l tional procedures. Fu rthe rr.o re, the switenes will t;E color radded to { ennance the functional control relationships. l t %%wd: N ftt it t uM y mag %. avk vm LL co f N /t t \\o 11 ru;WJ st. i

7 ~ _. Y ..g .r r SNPS-1 FSAR I.D.1 CONTROL r.00': DESIGN JEVIE'.iS 5rdR(_-Q]p Additional LILC0 Resoonse to NRC Findinq 6.20 - Containment Purae Control "alves The valves in question (LEG *A0V3SC, 3ED, 39C, and 39D) are 18" suppecssion cha:rber purge vaives. Prior to the decision to inert the Shorcham containc.ent, these valves were intended to purge the suppression chamber perioJically for hydrogen and pressure control. With the implementation of the contair:ent ir.erting system and the addition of a 6" vent line (IT46'A0V79A and D), the above valves will ng be uscd during normal opera:. ion. Valves 3CC, 3CD, 39C, and 39D will be ustd to parge the suppression chir;.ber prior to entry during shut down condition. Also, both valves hwe to be operated to perform the function and, from an operating point of vicu, relative switch /s. nimic location has little c: caning. The engineering and construction impact to relocate thc switches as suggested by the Cc.rnission is very significant and cannot '..e justified b since there is no bcnefit to changing frc't the present position. II.2 impact is significant due to the complex bccrier design within the panel to satisfy the scparation criteria for these and other controls on this panel. Each valve is individually key locked closed and under centrol of operatjonal procedures. Furthen:: ore, the switches vill be color padded to enhance the functional control relationships. In summary, Lilco believes that this individua' NRC finding does not have any real safety significance and givcn the aforementioned justifi-cation, piant r.:cdification is not warranted. 9 8/6/81

(- , _..b $.Y. O.. __ _.b c 7.10 On Panel 604, the relationship between the reset selector positions and the upscale and downscale trip indicators is not clear. (7.19) LILCO Rescense On Panel EC4 for the Off Gas Vent Pipe Radiation Monitors A and 8 and the Reactor Building Stancby Ventilation Radiation Monitor, a label will e added to indicate that the upscale /downscale pusn button must be depressec . to reset tne trip indications. T'1is wili De accomo'lisned crior to loading fuel. For the remaining nonitors en Panel 604, the reset switen is of the type sucn that onen the reset swi*cn is rotated either to the left or 2,. to tne rignt, the trip indicati0ns are reset. No A "'"e" ' I'F" 'ct'en m e-m m e,

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.m nes. . Omwsuvb '- kN h% MW Th7 =D ~ k LtLO ow MfL 52h IIRC Finding 7.19 On Panel 604, the relationship tetween the reset selector positions and the upscale and downscale trip indicatora is not clea r. (2) LILCO Response On panei 604 for the Off Can Vent Pipe Radiation :!onitors ALD and the Reactor ~ Y Building Standby Ventilation Radiation !!onitor, a label will be added to indicate that the upscale /downscale push button mus.t be depressed to reset the trip indications. This will be accompitsh(d prior to fuc1 load. For the remaining monitors on panci 604, the reset switch is of the type such that when the reset ~ ^ switch is rotated either to the.left or to the riant, the_ trip indications are reset. j SS f 4 No83 ) B.3 Labels on panel below recorders that protrude from the panel are hidden from view of standino coer.1 tors. 0 k o ,9 ) sg. g Ye } - wm q m s, x

I .._.c LILCO Restonsa Labels will te located on f.he recorder to provice clear readability to the operator. W M k... .M )

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Labels on panel below recorders that protrude from the panci are hidden fron view of standing, operators. Exanple: 11XP Panci and Panci 603 (1) LILCO Response Labels located below controls f, recorders util be reevaluated for adequacy prior to fuel load. Changes necessitated as a result of the reevaluation will be effected prior to fuel load. b k. 8.4 Label on reactor water cleanup return J-switen that states TO REACTOR VESSEL is misleading since the return flow path is through HPIC, RCIC, and feedwater lines to reach the reactor vessel. (9.4) LILCO Resoonse The return flow path of the reactor water cleanup system (WCU) to the vessel is through the feecwater system. The HPCI and RCIC flow paths also return to the vessel via the feecwater syste.w. The RWCU system coes not flow through the HPCI and RCIC systems to return to the vessel. ~ Operator classroom and field training enforces uncers.ancing of 'tne entire RWCU flow path. LILCO believes that a change in the labeling, as requested, would be misleading and detract fecm the prinry indication of , return ficw to the vessel. -Therefore - no further-ection-wiLI bg J.atan. 4 Cowd Lud h%w w mw\\ b) tno w owr 504M 3M.hM

SRC Findinc. 8.4 $N{( Q{ Label on reactor water cleanup return J - st. itch that states TO REACTOR VESSEL is misleading since the return flow path is through IIPCI, RCIC, and feedwater lines ' to reach the reactor vessel. (1) a-LlLCO Response y The return flow path of the reactor water cleanup syste (Rt.*CC) to the vessel is through the feedwater system. The IIPCI and RCIC fic.* paths also return to the vessel via the feedwater system. The PJ.!CP Systen does not ficw throu;h the llPCI and RCIC systens to return to the vessel. Operator classroca and field training, enforces understanding of the entire RUCL' flow puh. LILCO believes that a change in the labeling, as requested, would he nisleading and detract froa the primary indication of return flow to the vessel. ...... j. _..b k l\\._ ....L... - - h -, h -.. - - - / ~~ 8.3 There aro -o inst: uction tabeis for aparation af :no control transfer switches on the remote shutdown panel. (8.8) () . ~.. L:1.C0 Resconse Aopropriate procacures wi?'1 e ir.cated at the ren ta shatdowr: ;:ane l. Therefore, no instruction labels at the control transfer switchos are required. No-fuMhee 2

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( g,9 h 'A(A YM b%6 DN k t.N u> w o i S M t..-sTV i ~ ~ N Ok NRC Finding 8.8 3 %Q,,, Q 3 There are no instruction labels for operation of the control transfer switches on b the remote shutdown panel. (2) LlLCO Response Appropriate procedures will be located at the renate shutdown pincl. Therefore, no instruction labels at the coatrol transf er switches are required. l

1 .-.. -. - 0J-- / 9.2 Excessive referencing of printed detumentation located at the corputer control consolo is needed to interpret CRT displays. Information neeced to foterpret each CRT display sneuld be an integral part of the display. (9.2) LILOO Resnonse It is felt that suf ficient infe--ation presently exists on each graphic display for the cperator to inte' : ret the variable data (analog, digital. and calculatea valt'ss) and the tatkground portion of each display. Adcitionally, operations perscerel have reviewed each grachic display ar: their cen=ents were then incer:erstad into tha final design. Prior to loading fuel, ccerator training W11 be conducted for each grar' tic display in the areas of pu pose, use, c:rtent of data, and data interpretation. -No-eddition11_4.IL;0.e u luu wfUm7skerw r -L ~ DhumM b

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.8DMVR M Yi>d D.WLW kv L.WW E Ow S9LL-52T h W w NRC Findine, 9.2 5(Q(1 Q $ Q{ Excessive referencia;; of printed docu entation located at the computer contre; censole is needed to interpret CRT displays. Informtion needed to interpra t each CRT display should be an intergral part of the display. (2) 1.ILCO Response it is felt that sufficient information presently exists on each graphic displ r for the operator to interpret the variable data (analog, digital, and calc.: a t ed values) and the background portion o* each display. Additionally, operatic-personnel have reviewed each graphic display and their coamments were then incorporated into the final design. Prior to fuel load, Operator Training. t. be conducted for each graphic display in the arcan of purpose, use, content :i data and data interpretation. 11 %EIL h.h 9.6 The alarm printer print speed, (33 characters per second) is too slew :: ( provide timely printuuts whert a large numoer of alarms occur simultare:_s f or in apid succession. (9.15) LILCO :tesconse Printers will be upgraded to a rating of 200 characters per second f:- all alarm messanas.

A% Led 3 I 1 \\ Ssi R. No. \\.\\ 1.1 Construction of the supervisor's office is not complete. The acequacy f visual and voice contact between that office and the control reem could not be evaluated. (1.3) LILCO Rescense The Watch Engineer's office will be comp'eted prior to loading fuel. y '/ 4 t i u l and voice contact will ce verified. 11% N %d b WM lb W\\. L}L[o hs, pur h 5 TS R L - 52 5 NRC Finding ly 9 N k(- $$ 5 s Construct ton of the supervisor's of'fice is not complete. The adequacy of visual and voice contact between that office and the control rocr could not be eva lu a t ed. (1) LILCO Resp ~onse g The Watch Engineer's office will be conpleted prior to fuel load. 2.) SW R. A 1.3

1. 3 Tangled and kinkee sound-powered phone cords lying on the floor are a tripping hazard. Use of non-kink or self retracting cords is recommanded.

(1.5) LILCO Resoonse During normal operation, the use of sound powered phones in the control room is localized at specific panels. Phone cords will be of appropriate length and cose no trioning hazard.

o 4 VV\\ W h) klojYJi\\}M. M b [ thME h N ILL o /' ?RC Finding 9.15 SN Q( $@5 The alarm printer print speed, (30 characters per second) is too sinw to provide tinely printouts when a large nunber of alarms occur ninultaneously or in rapid succ e ss !.on. The recommended miniming printer speed is 30',1 Lines per ntnute, (200 characters per second at 40 charactern por line. (2) LILCC Response Printers will be upgraded to a rating; of 26hl characters per second for all alara nessages, as soon as feasibic, consistent with vendor availability. Nk .) ?. ' 'Jse Sf '!asnfr.g ;ei!ow ".49.?"

• .c 'nelie.ata '. hat :ata should be ignored s confusing.

(9.16) LILCO Resconse The use of flashing yellow "-99.9" to indicate an unknown analog value has been reviewed and approved by operations personnel. This convention is also consistent with G.E. Software. Multiple unknown analog values on most tiisplays during the audit was due to the fact that a large nusnber of analog points are currently awaiting checkout by startup personnel. During plant operation the occurrence of unknown values will be infrequent. No additional-t-il.CO-actitrrr vf f t-be-texen. Om %i ~ h4%3kStsshn Wrv] Mwsmk Lt W w % sw su mg..

- g m .-m. .,p = I , NRC Finding 9.16 h Q Q - { g{ of flashing yellou 99.9 to indicate that data should be ignored is confusing. LILCO Response b The use'of flashing yellow "- 29.9" to indicate an unkno.en analog value haa heen reviewed and approved by operations pers.onnel. T:iin convent ion is also consistent with C.E. sof t wa re. !!al t i ple-t"Wnown analog values on nort displays - durlag the audit was due to the fact that larna nuaber of analon points are currently ava t ting check out by startup personnel. During plant operation the. l occurence oL unknown values will be inirequent. L WM N i l D Owt uwvv; 5, i i k i t i

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