ML19254F057

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Summary of 790820 Meeting W/Util in Bethesda,Md Re Mod to Helium Circulator Auxiliary
ML19254F057
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 10/20/1979
From: Kuzmycz G, Williams P
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 7911050528
Download: ML19254F057 (98)
v • d • e


Text

i, DISTRIBUTION Docket File R. DeYoung NRC PDR W. Kreger R. Houston 2)LocalPDR

UdL, J. Collins NRR Reading V. Moore H. Denton M. Ernst E. G. Case R. Denise D. Vassallo T. Speis D. Eisenhut S. Varga R. Vollmer D. Ross D. Skovholt P. Check W. Haass G. Lainas P. Collins B. Grimes J. Stolz J. Miller R. Baer W. Gammill
0. Parr i

ACRS (16)

C. Heltemes SD(7)

R. Mattson H. Ornstein F. Schroeder R. Meyer J. Knight M. Tokar R. Bosnak F. Litton S. Pawlicki L. Phillips

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F. Schauer T. Dunning R. Tedesco W. Paulson T. Novak R. Ireland K. Kniel R. Schamberger Z. Rosztoczy-R. Woodruff, I&E W. Butler G. Madsen, I&E, Reg. IV S. Hanauer P. Williams V. Benaroya Project M? nager - G. Kuzmycz R. Satterfield

'ttorney, ELD F. Rosa i-E. Hylton e

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NUCLEAR REGULATORY COMMISSION g

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.E WASHING ton, D. C. 20555 QKX]

OCT 2 0 M i

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50-267 l

LICENSEE:

Public Service Company of Colorado (PSCo)

FACILITY:

Fort St. Vrain (FSV)

SUBJECT:

SUMMARY

OF MEETINGS HELD ON AUGUST 20, 1979 TO DISCUSS I

FSV ITEMS 6

A meeting was held in the Bethesda offices of the Nuclear Regulatory Connission on August 20, IM9 to discuss various items related to FSV operation. A list of meeting attendees is presented in Enclosure 1.

The meeting agenda along with viewgraphs presented by PSCo are presented in Enclosure 2.

The purpose of the meeting was to give PSCo the opportur,i'.y of presenting an update of their work dealing with the items on the agenda (see Enclosure 2).

All of the items covered will be discussed in subsequent progress meetings.

I.

Highlights of Proposed ISI Plan 1.

The modified ISI provisions were proposed in a document entitled, " Progress Report on the Program to Review Fort St. Vrain In-service Inspection and Testing, Preliminary Development Phase." The document contained proposals on score and philosophy, criteria, definitions, equipment clarification, general rules and requirements for inspecting and testing classified equipment.

Illustrations were provided of proposed requirements for selected plant systems. PSCo is seeking our generalized approval for the outlined approach and aims to have the full plan ready for implementation in January 1980.

Implementation would be by modifications or additions to the Technical Specifications.

2.

The PSCo plan would be customized to the Fort St. Vrain safety system design but make use of the draft ASME Code (i.e.,Section XI, Division 2,1977) where PSCo judges it appropriate. The Fort St. Vrain ISI provision would be developed by PSCo under a philosophy that, "... recognizes the generic difference betw6 m light water cooled reactors and the Fort St. Vrain High Temperature Gas Cooled Reactor, particularly those difference which provide an inherently superior level of safety." The basic approach would be to check the condition of safety related items on a regular basis to a standard of assurance commensurate with the importance of the safety function

, of the item.

6 3.

The general surveillance requirements include a statement which would exempt pressure boundaries, both primary and secondary, from both surface and l

volumetric NDT surveillance.

4.

Inspection and testing would be verified by the PSCo quality assurance I

program and indepently verified by the OIE Region IV. Only those portions of the plant built soecifically to the ASME Code would be subject to inspection by qualified inspectors from qualified agencies.

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OCT 40 mg 5.

No systems related to reactor safety were evidently built to ASME Code Those portions of the plant which were built to ASME Code are inspecteif under the proversions of Section 8.

At the August 20 meeting PSCo agreed to provide a list of equipment in this category.

PSCo will also identify inspections and testing for safety related equipment that is not now currently included in the Technical Specifications.

6.

The general provisions would provide for replacements using construction code requirements based on standards consistent with those in effect during design and construction. For example, the helium circulator auxiliary system which is to be extensively modified to provide greater separation to severe each loop will be built to B 31.1 rather than to ASME Section III.

7.

Four surveillance classifications are proposed based on safety function.

Surveillance Class I would 7ertain to equipment necessary & cope with Design Basis Accident No. 1 (i.e., the sustained loss of forced circulation).

Class II would pertain to all other design basis accidents. Class III includes safety related structeres, pressure boundaries, and auxiliary systems which handle and store spent fuel. Class IV includes other systems which provided additional protection such as fire protection and radioactive wastes.

8.

T.%e ASME Code would be followed for Class I pumps and valves and PCRV structural integrity will follow Article 1GK of the ASME Code.

In this latter case the helium permeation test (Tech. Spec. SR-5.2.13) would be considered an alternate to the code required 100 percent inspection of the top and bottom needs.

Among(a)e many new surveillance Technical Specifications to be proposedisolation v 9.

th are:

an operability check on the loop ments for operation and leak checks on the PCRV isolation valves, and (c) inspectic of bolting of the PCRV cover plates. These items were stated as priority items for attention in earlier comunications with PSCo.

10. No report wad made on PGX graphite surveillance as information is still being developed in this area.

II. Modifications to Helium Circulator Auxiliaries Modifications to the systems supplying buffer helium to the bearings of the helium circulators were described. The objectives of these modifications are: (1) minimize the possibility of water entering the buffer heliem path, (2) improve control of low pressure separator water level under upset conditions, (3) minimize the possibility of abnomal high helium pressure in the low pressure separator and at the helium recovery compressor suctions, and (4) minimize and control the release of radioactive cas and water into the reactor building. The specific modifications to achieve each objective were described. We stated that these modifications appeared consirtent with achieving the stated objectives. We requested that PSCo.

femally submit a description of the forthcoming equipment and system changes.

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OCT e 01979 e III. Decalibration of Ex-Core Detectors Neutron flux measured by the six detectors on the outer surface of the PCRV requires calibration against a heat balance at intervals of 10 percent power change. Electronic circuitry was proposed that would automatically provide this calibration and adjust the setting of the trip point pertaining to protection against the rod withdrawal accident. The reactor would also trip on rapid rates of neutron flux increase as seen by the ex-core detectors.

In our discussions with PSCo we stated we were satisfied with the physics aspects of the proposed approach and that our next step would be to review the circuitry and other electrical matters.

IV.

PCRV Liner Hotspots An undate of the December 1977 report on PCRV liner hotspots was presented.

Based on data obtained above 40 percent power only the top head penetrations present, a present concern.

In regions 5, 34 and 35 themocouples were located that indicated ambient temperatures of 1800F.

This is the temperatures limit at which the control rod drive system was tested.

Weekly scram operability tests have been increased from monthly to weekly.

Further infomation on this situation will be forthcoming from PSCo.

V.

Moisture Monitor Circuitry Modifications M;difications to the moisture monitor circuitry were proposed to make permanent time delay circuitry to prevent re-occurrence of the two-loop dump event. Circuit changes were also described that would reduce spruious moisture monitor trips.

PSCo was advised to formally submit a description of these circuit changes for our review.

VI.

F_uel Element Thermal Stress during Fire-Water Cooldown Themal stresses were calculated in a fuel element for gas-graphite 6

temperature differences of 15000F and 1000. A maximum themal stress of 470 PSI (tensile) was found for the 15000F case. We infomed PSCo that our consultant at LASL also found that thermal stress in the fuel would be inconsequential and was now concentrating on the core support blocks.

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- h ^ eter. Williams George Kuzmyc:

h Division of Project Management 19

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ENCLOSURE 1 MEETING ATTENDEES NRC PSCo G. Kuzmycz F. Swart P. Williams J. Gahm R. Ireland J. Reesy M. Tokar F. Lesko W. Brooks M. Padovan F. Tilson M. Holmes '6[

m., secar4C Pusuc Stnyict Courany or Cotonaco cm Dd-1L 3 ^ To: Mr. Georce Kuzmycz ENCLOSURE 2 Advanced Reactors U.S.N.R.C. Frm: H.H. Holmes M-Public Servico Crupany of Colorado PROPOSED PSC/NRC HEETING AGENDA Date: August 20-21, 1979 I I. InserviceInsrection(ISI) 1) Review of proposed program md systera classifications 2) Review cf pr'oposed surveillance plan 3) Review of proposed surveillance of selected plant systems a) Reserve Shutdown Systs b) Reactor Building Ventilation Exhaust System c) Reactor Building Overpressure Protection System d) Purificat!9n Cooling Water System e) FCRV Safety Valves f) PCRV 4) Evaluation of previch HRC ISI concerns 5) Discussion of open items fra Itay 2,1979 Meeting Hinutes SV _II. Helium Circ 91ator Auxiliary System Hodifications it .III.. Deca 11bration of Prestressed Concrete Reactor Vessel (PCRV) Neutron Detectors - Physics Aspects and Circuit Changes D .. I V. PCRV Liner (CRD Penetration) Hot Spots - Update on December 1977 Report fa V. Hoisture Monitor Circuitry Modifications M VI. Fuel Elements - Stress Encountered AfterLOFC Accident with Cooling Initiated on Firewater. VII. Other Items rG ' Ul3P! - }>'" ' r,

PROGRESS REPORT ON THE PROGRAM TO REVIEW FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PRELIMINARY DEVELOPMENT PHASE s J ) E '4

PRELIMINARY DEVELOPMENT PHASE PROGRESS REPORT 0 PROPOSED PROGRAM SCOPE AND PHILOSOPHY. O PROPOSED CRITERIA, DEFINITIONS AND ASSIGNMENTS FOR CLASSIFYING EQUIPMENT IMPORTANT TO SAFETY. O PROPOSED GENERAL RULES AND REQUIREt:ENTS FOR INSPECTING AND TESTING EQUIPMENT ASSIGNED TO THE VARIOUS CLASSIFICATIONS. O PROPOSED SURVEILLANCE REQUIREMENTS FOR SELECTED PLANT SYSTEMS - THE RESULTS OF APPLYING THE GENERAL RULES AND REQUIREMENTS DURING THE REVIEW PROCESS FOR THESE SYSTEMS. M N J ) .71 t O

EROPOSED FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PROGRAM PURPOSE - THE OVERALL PURPOSE OF THE INSERVICE INSPECTION AND TESTING PROGRAM AT FORT ST. VRAIN IS TO PROVIDE A CONTINUING ASSURANCE THAT COMPONENTS, SYSTEMS, AND STRUCTURES IMPORTANT TO SAFETY, ARE SAFE AND ARE AVAILABLE TO PERFORM THEIR SAFETY FUNCTIONS WHEN REQUIRED. THE PROGRAM WILL ESTABLISH SPECIFIC SURVEILLANCE RULES AND REQUIREMENTS IN ORDER TO: O GENERATE OBJECTIVE EVIDENCE ON A CONTINUING BASIS THAT AN ADEQUATE LEVEL OF NilCLEAR POWER PLANT SAFETY IS BEING MAINTAINED. O PROVIDE EARLY DETECTION OF DETERIORATION / DEGRADATION OF SAFETY-RELATED FACTORS FROM BASELINE CONDITIONS. O PROVIDE ADDITIONAL ASSURANCE THAT THE PLANT IS BEING MAINTAINED ] IN A SAFE CONDITION THROUGHOUT ITS LIFETIME. J 3 (N

EBlLOSOPHY - THE DEVELOPt"NT OF RULES AND REQUIREMENTS FOR SURVEILLANCE OF EQUIPMENT IMPORTANT TO SAFETY AT FORT ST. VRAIN WILL BE BASED ON A PHILOSOPHY WHICH CONSIDERS THE CONCERNS FOR OPERATIONAL READINESS OF SAFETY SYSTEMS AS WELL AS THE CON-CERNS FOR INTEGRITY OF THE REACTOR COOLANT PRESSURE BOUNDARY. THIS PHILOSOPHY RECOGNIZES THE IMPORTANCE OF VERIFYING PRESSURE BOUNDARY INTEGRITY TO PREVENT THE OCCURANCE OF A FISSION PRODUCT RELEASE ACCIDENT; BUT IN ADDITION, IT ALSO RECOGNIZES THE IMPORTANCE OF BEING ABLE TO MITIGATE THE CONSEQUENCES 0F AN EVENT WHICH COULD ADVERSLY AFFECT PUBLIC HEALTH AND SAFETY. FURTHERMORE, THE PHILOSOPHY RECOGNIZES THE GENERIC DIFFERENCES BETWEEN LIGHT WATER COOLED REACTORS AND THE FORT ST. VRAIN HIGH TEMPERATURE GAS COOLED REACTOR, PARTI-CULARLY THOSE DIFFERENCES WHICH PROVIDE AN INHERENTLY SUPERIOR LEVEL OF SAFETY. THE BASIC APPROACH TO BE USED IN DEVELOPING THE RULES FOR FORT ST. VRAIN IS TO ESTABLISH A PROGRAM THAT WILL CHECK THE CONDITION OF PLANT STRUCTURES, SYSTEMS, AND COMPONENTS ON A REGULAR BASIS TO A STANDARD OF ASSURANCE COMMENSURATE WITH THE IMPORTANCE OF THE SAFETY FUNCTION OF THE ITEM. CLASSIFICATIONS WILL BE ESTABLISHED TO PROVIDE THESE LEVELS OF IMPORTANCE AND APPROPRIATE SURVEILLANCE REQUIREMENTS (I.E. TESTS, EXAMINATIONS, OBSERVATIONS, MONITORING, ETC.) 3 SPECIFIED FOR EACH CLASSIFICATION. THE SURVEILLANCE REQUIREMENTS WILL BE ] BASED ON, DEVELOPED FROM, AND ENHANCE THE EXISTING SURVEILLANCE REQUIREMENTS CONTAINED IN THE FORT ST. VRAIN TECHNICAL SPECIFICATIONS. r, '3 M

SCOPE - THE INSERVICE INSPECTION AND TESTING PROGRAM PROVIDES RULES AND REQUIREMENTS FOR SURVEILLANCE OF PLANT STRUCTURES, SYSTEMS, AND COMPONENTS AT THE FORT ST. VRAIN NUCLEAR GENERATING STATION. THE PROGRAM APPLIES TO THOSE PLANT EQUIPMENT ITEMS WHICH, IF FAILED, COULD INTERFERE WITH REACTOR SHUTDOWN, INTERFERE WITH REMOVAL OF DECAi HEAT OR RELEASE EXCESSIVE RADIOACTIVITY TO THE ENVIRON-MENT. SUCH ITEMS ARE GENERALLY CLASSIFIED AS " SAFETY-RELATED" AND INCLUDE EQUIPMENT DESIGNATED "FSAR SAFETY CLASS I", " SAFE SHUTDOWN", " ENGINEERED SAFEGUARDS", " STANDBY POWER", " EMERGENCY COOLING", AND " BACKUP EMERGENCY COOLING". THE PROGRAM ALSO APPLIES TO OTHER PLANT EQUIPMENT ITEMS WHICH, IF UNAVAILABLE MAY REDUCE OR DEGRADE THE MARGIN OF SAFETY THEY PROVIDE. THE PROGRAM FURTHER CLASSIFIES SAFETY-RELATED AND OTHER EQUIPMENT ACCORDING TO THEIR IMPORTANCE TO PLANT SAFETY AND PROVIDES THAT EACH EQUIPMENT ITEM BE PERIODI-CALLY TESTED, EXAMINED, OBSERVED, CALIBRATED, OR MONITORED AS APPROPRIATE. THE PROGRAM PROVIDES CRITERIA FOR ACCEPTANCE OF RESULTS, REPAIR AND REPLACEMENT, AND PERSONNEL QUALIFICATI0t'1; IT ALSO DEFINES RESPONSIBILITIES, EXAMIWATION TECHNIQUES AND RECORD KEEPING / REPORTING REQUIREMENTS. t ' s J J \\ (T}

9 -i i FSAR SAFETY SAFE SHUTDOWN CLASS I EQUIPMENT COOLING EQUIPMENT ~ FSAR SAFETY CLASS II FSAR SAFETY CLASS II. EQUIPMENT EQUIPMENT g ....4 4 ..-_,6 7 ,1. o f \\ . + { .\\ i ~ ENGINEERED SAFEGUARDS [ [ h BACKUP EMERGENCY (REACTOR BUILDING COOLING EQUIPMENT EXHAUST) (ACM SYSTEM) / \\ \\ .4 , r -- - --ACTIVITY / RADIATION REACTIVITY DECAY llEAT

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- ' ~ ' CONTROL CONTROL ' - ' REMOVAL ~T~"'-'-~~ j.. -. ..j. f SAFETY FUNCTIONS' _i i f... g......._ j l SAFETY-RELATED t i ~ " EQUIPMENT DESIGNATED FOR SURVEILLANCE ~~~! ~ ~ ~~~ I '~~-~ ~ I a t i i. { _r._.. l I I e 1 4 i ' g 3 g e . i l J s -1 1 J (y.

t e PROPOSED SURVEILLANCE PLAN GENERAL SURVEILLANCE REQUIREMENTS O SCOPE AND RESPONSIBILITY - OVERALL PROGRAM WILL BE BASED ON EXISTING TECHNICAL SPECIFICATION SURVEILLANCE PROGRAM - PLANT EQUIPMENT IMPORTANT TO SAFETY WILL BE CLASSIFIED TO PROVIDE A BASIS FOR SURVEILLANCE INSPECTION AND TEST REQUIREMENTS m s J <u 47

a GENERAL SURVEILLANCE REQUIREMENTS (CONTINUED) 0 INSPECTION AND TESTING - OPERATIONAL READINESS AND STRUCTURAL INTEGRITY OF PLANT EQUIPMENT WILL BE DETERMINED BY SURVEILLANCE TESTING AND SURVEILLANCE INSPECTIONS. - SURVElli ANCE TESTING WILL CONTINUE TO BE EMPHASIZED AND WILL INCLUDE OP: RATIONAL TESTS, FUNCTIONAL TESTS, PRESSURE TESTS, LEAKAGE Tf3TS, AND CALIBRATION TESTS. - SURVEfi, )CE INSPECTIONS WILL BE LIMITED TO VISUAL EXAMINATIONS, OBSCRVATIONS, AND MEASUREMENTS. - SURFACE AND VOLUMETRIC EXAMINATIONS USING NON-DESTRUCT'IVE TEST METHODS WILL BE USED ONLY FOR INVESTIGATIVE PURPOSES - NOT FOR SURVEILLANCE. - SPECIFIC SURVEILLANCE INSPECTION AND TEST REQUIREMENTS WILL CONTINUE TO BE SPECIFIED IN THE TECHNICAL SPECIFICA-TIONS ISSUED BY THE NRC AS PART OF THE OPERATING LICENSE 3 PROVISIONS. s J

INSPECTION AND TESTillQ (CONTINUED) - SURVEILLANCE INSPECTION AND TEST INTERVALS WILL CONTINUE TO BE SPECIFIED FOR EACH SPECIFIC SURVEILLANCE REQUIREMENT IN THE TECHNICAL SPECIFICATIONS. - QUALIFICATION REQUIREMENTS FOR INSPECTORS WILL CONTINUE TO BE INCLUDED IN THE QUALITY ASSURANCE PROGRAM. - INSPECTION AND TESTING WILL CONTINUE TO BE VERIFIED BY THE QUALITY ASSURANCE DEPARTMENT. - INSPECTION AND TESTING WILL CONTINUE TO BE INDEPENDENTLY VERIFIED BY THE NRC DIVISION OF INSPECTION AND ENE0RCEMENT (REGION IV). ..M s J .~ t,

GENERAL SURVEILLANCE REQUIREMENTS (CONTINUED) 0 ACCEPTANCE STANDARDS - ACCEPTANCE STANDARDS WILL CONTINUE TO BE SPECIFIED FOR EACH SPECIFIC INSPECTION OR TEST REQUIREMENT. FOR THOSE INSTANCES WHEN ACCEPTANCE STANDARDS CANNOT BE SPECIFIED, AN ENGINEERING EVALUATION OF THE IN-SPECTION OR TEST RESULTS WILL BE REQUIRED TO DETERMINE ACCEPTABILITY. O REPAIR PROCEDURES - GENERAL REQUIREMENTS FOR COMPONENT PRESSURE BOUNDRY REPAIRS WILL BE DEVELOPED FOR INCLUSION IN THE PROGRAM. ~ 0 REPLACEMENTS - CONSTRUCT. ION CODE REQUIREMENTS WILL CONTINUE TO BE BASED ON STANDARDS CONSISTENT WITH THOSE IN EFFECT DURING DESIGN AND CONSTRULTION. - VERIFICATION OF ACCEPTABILITY WILL CONTINUE TO BE BASED ON SUITABILITY EVALUATIONS AS REQUIRED BY THE PSC QUALITY ASSURANCE DEPARTMENT. - GENERAL REQUIREMENTS FOR REPLACEMENTS WILL BE DEVELCPED FOR INCLUSION IN THE PROGRAM. a 4 ~ un

SURVEILLANCE CLASS III ACTIVITY / RADIATION CONTROL EQUIPMENT - PCRV STRUCTURE (11) - PCRV LINER (11) - PCRV PENETRATIONS AND CLOSURES (11) - PCRV THERMAL BARRIER (ll) - PCRV INTERNAL STRUCTURE (CORE SUPPORT) (11) - PCRV AUXILI ARY SYSTEM (11) * - FUEL HANDLING MACHINE (13) - FUEL STORAGE FACILITY (14) - STEAM GENERATOR (22) SPENT FUEL COOLING EQUIPMENT - FUEL STORAGE FACILITY AUXILI ARY SYSTEM (14) *

  • ONLY THOSE PORTIONS MARKED AS SAFETY-RELATED

'( ON THE SR-6-1 DRAWINGS J J L.) t I-

S11RVEILLANCE CLASS IV INSTRUMENTATION AND CONTROL EQUIPMENT - OVERALL PLANT CONTROL SYSTEM (93)* - CONTROL R0D AND ORIFICING C81 SYSTEM (93)* - PCRV INSTRUMENTS AND DATA ACQUISITION SYSTEM (93)* - FAST GAS AND IODINE SAMPLING SYSTEM (93)* - Ss T/C AND S/G INSTRUMENTATION SYSTEM (93)* - ANALYTICAL INSTRUMENTATION SYSTEM (93)* - AREA RADIATION MONITORING SYSTEM (93)* - COOLANT MEASUREMENT DISPLAY SYSTEM (93)* - SEISMIC INSTRUMENTATION SYSTEM (89) - PCRV SURFACE TEMPERATURE MONITORS (73) - PCRV AREA zip AND TEMPERATURE MONITORS 173)

  • 0NLY THOSE PORTIONS

- CONTROL ROOM TEMPERATURE CONTROLLER (75) WHICH PROVIDE AN ADDI- - 480V ROOM TEMPERATURE INDICATOR (75) TIONAL SAFETY FUNCTION - AIR EJECTOR VENT MONITOR (41) - HOT REHEAT PIPING MONITORS (22) - HOT REHEAT PIPING POWER RELIEFS (52) a s J h

SURVEILLANCE CLASS IV (CONTINUED) RADIOACTIVE WASTE MANAGEMENT EQUIPMENT - RADICACTIVE LIQUID WASTE STORAGE SYSTEM (62) - RADIOACTIVE GAS SYSTEM (63) EJRE PROTECTION EQUIPtiERI. - FIRE WATER SYSTEM - DELUGES / SPRAYS / HOSE REELS (45) - CONTROL COMPLEX FIRE DETECTION AND ALARM SYSTEM (XX) - CONTROL COMPLEX HALON SYSTEM (45) - CONTROL COMPLEX EMERGENCY BREATHABLE AIR SYSTEM (45) - CONTROL COMPLEX FIXED WATER SPRAY SYSTEM (45) - CONTROL COMPLEX CABLE COATING AND 3EAL SYSTEM (XX) -C 2 SYSTEM (51) -A J D iJ-D'

SURVEILLANCE INSPECTION ANL TEST REQlilREMENTS A. OPERATIONAL READINESS 1. SYSTEM TESTIME O SURVEILLANCE CLASS I - THE OPERATIONAL READINESS OF SYSTEMS ASSIGNED TO SURVEILLANCE CLASS I SHALL BE DEMONSTRATED BY. SURVEILLANCE TESTING AND MONITORING. TESTING AND MONITORING REQUIREMENTS SHALL BE BASED ON VERIFYING OVERALL SYSTEM PER-FORMANCE AND AVAILABILITY OF EMERGENCY POWER SOURCES. 0 SURVEILLANCE CLASS II - THE OPERATIONAL READINESS OF SYSTEMS ASSIGNED TO SURVEILLANCE CLASS II SHALL BE DEMONSTRATED BY NORMAL OPERATION OF THE SYSTEM. O SURVEILLANCE CLASS III AND IV - SYSTEM TESTING IS NOT APPLICABLE. s 1 D t -4 et

A.' OPERATIONAL READINESS (CONTINUED) 2. COMPONENT TESTING 0 PUMPS - SURVEILLANCE CLASS I - THE OPERATIONAL READINESS OF PUMPS ASSIGNED TO SURVEILLANCE CLASS I SHALL BE DEMONSTRATED ONCE EACH QUARTER BY SURVEILLANCE TESTING. THE RULES AND RE-QUIREMENTS FOR SURVEILLANCE TESTING OF THESE PUMPS WILL BE BASED ON ARTICLE IGP OF THE DRAFT ASME CODE. - SURVEILLANCE CLASS II - THE OPERATIONAL READINESS OF PUMPS ASSIGNED TO SURVEILLANCE CLASS II SHALL BE DEMONSTRATED BY NORMAL SYSTEM OPERATION OR BY SURVETLLANCE TESTING TO EXERCISE THOSE PUI4PS NOT NORMALLY IN OPERATION ONCE EACH QUARTER. - SURVEILLANCE CLASS III AND IV - THERE ARE NO PUMPS ASSIGNED

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TO SURVEILLANCE CLASSES III AND IV. s 3) _w [ b

2. COMPONENT TESTING (CONTINUED) O COMPRESSORS - SURVEILLANCE CLASS I,

III,

- THERE ARE NO COMPRESSORS ASSIGNED TO SURVEILLANCE CLASSES I AND III. - SURVEILLANCE CLASS II AND IV - THE OPERATIONAL READINESS OF COM-PRESSORS ASSIGNED TO CURVEILLANCE CLASSES II AND IV SHALL BE DEMON-STRATED BY NORMAL OPERATION OR BY SURVEILLANCE TESTING TO EXERCISE THOSE COMPRESSORS NOT NORMALLY IN OPERATION ONCE EACH QUARTER.- THE OPERATIONAL READINESS OF COMPRESSOR AUZILIARY DRIVES ASSIGNED TO SURVEILLANCE CLASS II SHALL BE DEMONSTRATED BY SURVEILLANCE TESTING ONCE EVERY TWO YEARS. s J .) D L 't V

2. COMPONENT TESTING (CONTINUED) 0 VALVES - SURVEILLANCE CLASS I - THE OPERATIONAL READINESS OF VALVES ASSIGNED TO SURVEILLANCE CLASS I SHALL BE DEMONSTRATED BY SURVEILLANCE TESTING. THE RULES AND REQUIREMENTS FOR SURVEILLANCE TESTING OF THESE VALVES WILL BE BASED ON ARTICLE IGV 0F THE DRAFT ASME CODE. - SURVEILLANCE CLASS II, III. AND IV - THE OPERATIONAL READI-NESS OF VALVES ASSIGNED TO SURVEILLANCE CLASSES II, III, AND IV SHALL BE DEMONSTRATED BY NORMAL OPERATION OR BY SURVEILLANCE TESTING TO EXERCISE THOSE VALVES WHICH DO NOT NORMALLY OPERATE. ISOLATION VALVES SHALL BE TESTED AT LEAST ONCE EACH YEAR AND SAFETY / RELIEF VALVES SHALL BE TESTED AT LEAST ONCE EVERY FIVE YEARS. 3. INSTRUMENTATION AND CONTROL TESTING O THE OPERATIONAL READINESS OF INSTRUMENTATION AND CONTROL CIRCUITS J SHALL BE DEMONSTRATED BY SURVEILLANCE TESTING. GENERALLY, INSTRU- ) MENT ACCURACY SHALL BE DEMONSTRATED ONCE EACH YEAR BY A CALIBRATION TEST. APPLICABLE INDICATION, ALARM, CONTROL, AND PROTECTIVE CIRCUIT 'o PERFORMANCE SHALL BE DEMONSTRATED BY A FUNCTIONAL TEST AT LEAST ONCE EACH YEAR DEPENDING ON THE IMPORTANCE OF THE SAFETY FUNCTION. c.

B.- STRUCTURAL INTEGRITY 1. EBESSURE BOUNDRY LEAKAGE TESTING 0 SYSTEM LEAKAGE TESTING - STRUCTURAL INTEGRIlY OF ALL ACCESSIBLE SAFETY RELATED PIPING SYSTEM PRESSURE BOUNDRIES SHALL BE VERIFIED BY EXAMINATION FOR LEAKAGE ONCE EACH YEAR WHEN THE SYSTEM IS AT OR NEAR NORMAL WORKING PRESSURE. O PCRV PENETRATION LEAKAGE TESTING - STRUCTURAL INTEGRITY OF PCRV PENETRATION PRESSURE BOUNDRIES SHALL BE VERIFIED BY LEAKAGE MONITORING. PENETRATION CLOSURE SEAL LEAKAGE SHALL BE MONITORED, CONTINUOUSLY FOR ALL PENETRATION GROUPS AND MEASURED ONCE EACH MONTH FOR EACH PENETRATION GROUP. m s J AD d

B. STRUCTURAL INTEGRITY (CONTINUED) 2. STRUCTURE INSPECTIONS 0 ECRM - STRUCTURAL INTEGRITY OF THE PCRV SHALL BE VERIFIED BY ~ SURVEILLANCE INSPECTIONS. THESE INSPECTIONS SHALL CONSIST OF VISUAL EXAMINATIONS, MONITORING, MEASUREMENTS, AND INSPECTION OF SPECIMENS. THE RULES AND REQUIREMENTS.FOR THESE INSPECTIONS WILL BE BASED ON ARTICLE IGK OF THE DRAFT ASME CODE. O SIEAM GENERATORS - STRUCTURAL INTEGRITY OF THE STEAM GENERATORS SHALL BE VERIFIED BY CONTINUOUSLY MONITORING AND MAINTAINING PRIMARY COOLANT MOISTURE, REHEAT STEAM ACTIVITY AND FEEDWATER CHEMISTRY BELOW ACCEPTABLE LIMITS. O REACTOR INTERNALS - STRUCTURAL INTEGRITY OF THE SUPPORT STRUCTURES FOR THE REACTOR CORE OR THE PCRV THERMAL BARRIER SHALL BE VERIFIED, WiiERE FEASIBLE, BY SURVEILLANCE INSPECTIONS OF MATERIAL SPECIMENS WHICH HAVE BEEN EXPOSED TO CONDITIONS SIMILAR TO THE SUPPORT COMPO-NENTS. THE FREQUENCY OF SUCH INSPECTIONS SHALL BE BASED ON EVALUATIONS i 0F THE MATERIAL PROPERTIES. J .) 23 L3 t 9

PROPOSED SURVEILIANCE REQUIREMENTS RESERVE SHUTDOWN SYSTEM A. OPERATIONAL READINESS 1. SYSTEM TESTING / MONITORING A) CONTINUE EXISTING SR 5.1.2A WHICH REQUIRES EACH OF THE SHUTDOWN HOPPERS TO BE PRESSURIZED 10 PSI ABOVE REACTOR PRESSURE ONCE EACH QUARTER TO DEMONSTRATE THAT THE HOPPER PRESSURIZING LINES ARE CLEAR, THAT THE HOPPER PRESSURIZING VALVES ARE OPERABLE, AND THAT THE HOPPER RUPTURE DISCS ARE INTACT. B) THE HELIUM STORAGE BOTTLE PRESSURES ARE MONITORED CONTINUOUSLY AND ALARMED BELOW 1500 PSIG TO ENSURE THAT THE REQUIRED ACTUATINL 'RESSURE IS MAINTAINED FOR THE RESERVE SHUTDOWN HOPPERS AND THEREFORE THAT THEY ARE OPERABLE.A NEW SR IS NOT REQUIRED TO DEMONSTRATE THAT THE SYSTEM IS CAPABLE OF FUNCTION-ING PROPERLY. C) ADD A NEW SR WHICH REQUIRES THE NITROGEN STORAGE BOTTLE PRESSURES 1 TO BE MONITORED ONCE EACH WEEK TO ENSURE THAT THE REQUIRED ACTUATING PRESSURE IS MAINTAINED FOR THE HOPPER PRESSURIZING VALVE AIR HEADER 3] DURING ACM SYSTEM OPERATION. )

A. OPERATIONAL READINESS (CONTINUED) 2. COMPONENT TESTING A) RESERVE SHUTDOWN HOPPERS - CONTINUE EXISTING SR 5.1.2C WHICH REQUIRES ONE RESERVE SHUTDOWN HOPPER TO BE TESTED IN THE HOT SERVICE FACILITY AT EACH OF THE FIRST FIVE REFUELING SHUTDOWNS AND THEN AT EVERY OTHER REFUELING SHUTDOWN TO DEMONSTRA.'E THAT THE RUPTURE DISC BURSTS AT THE CORRECT SET PRESSURE AMD THAT THE POISON MATERIAL RELEASES PROPERLY FROM THE HOPPER AND HAS N01 DETERIORATED. B) HOPPER PRESSURIZING VALVES - ADD A NEW SR WHICH REQUIRES EACH OF THE AIR-OPERATED HOPPER PRESSURIZING VALVES TO BE EXERCISED AT EACH RE-FUELING SHUTDOWN TO DEMONSTRATE THAT THE VALVE OPERATOR STROKES PROPERLY. C) PENETRATION CHECK VALVES - ADD A NEW SR WHICH REQUIRES EACH OF THE HOPPER PRESSURIZING LINE CHECK VALVES TO BE TESTED EACH TIME THE PENETRATION IT IS CONNECTED TO IS OPENED FOR REFUELING TO DEMONSTRATE THE ABILITY OF THE CHECK VALVE TO FUNCTION PROPERLY IN THE REVERSE DIRECTION. D) ACM OUICK-DIS'ONNECT VALVES - ADD A NEW SR WHICH REQUIRES EACH OF THE ACM OUICK-DISCONNECT VALVES TO BE CONNECTED AND DISCONNECTED j ONCE EACH QUARTER TO DEMONSTRATE THAT THIS VALVE IS CAPABLE OF ta FUNCTIONING PROPERLY. r. I

A.- OPERATIONAL READINESS (CONTINUED) 3. INSTRUMENTATION AND CONTROL TESTING A) RESERVE SHUTDOWN HOPPER - CONTINUE SR 5.1.2A WHICH REQUIRES THE RESERVE SHUTDOWN HOPPER PRESSURE ALARM CIRCUIT TO BE TESTED ONCE EACH QUARTER IN CONJUNCTION WITH THE SYSTEM TEST TO DEMONSTRATE THAT THE PRESSURE SWITCH AND ALARM FUNCTION PROPERLY. CONTINUE SR 5.1.2E WHICH REQUIRES THE RESERVE SHUTDGWN HOPPER PRESSURE SWITCH TO BE CALIBRATED WHENEVER THE CONTROL AND ORIFICE ASSEMBLY IS REMOVED FROM THE REACTOR FOR MAIN-TENANCE TO DEMONSTRATE THAT THE PRESSURE SWITCH OPERATES AT THE CORRECT SET PRESSURE. B) NORMAL PRESSURIZING LINE - CONTINUE SR 5.1.2D WHICH REQUIRES THE-PRESSURIZING LINE PRESSURE ALARM CIRCUIT TO BE TESTED ONCE EACH QUARTER IN CONJUNCTION WITH THE SYSTEM TEST AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE SWITCH AND ALARM CIRCUIT FUNCTIONS PROPERLY. C) TEST PRESSURIZING LIRE - CONTINUE SR 5.1.2B WHICH REQUIRES THE TEST GAS PRESSURIZING LINE PRESSURE INDICATION AND ALARM CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE INDICATOR, THE PRESSURE INDICATING SWITCH, AND THE ALARM CIRCUIT FUNCTION PROPERLY. m } D) PRESSURIZING VALVES - MODIFY EXISTING SR 5.1.2A TO REQUIRE THE PRES-SURIZING VALVE CONTROL CIRCUIT EE TESTED ONCE EACH QUARTER TO DEMON-STRATE CONTINUITY BETWEEN THE CONTROL ROOM AND THE LOCAL CONTROL PANEL. (NOTE: THIS TEST IS PRESENTLY PERFORMED). ADD A NEW SR WHICH REQUIRES THE PRESSURIZING VALVE CONTROL CIRCUIT TO BE TESTED AT EACH REFUELING SHUTDOWN TO DEMONSTRATE THAT THE CONTROL ROOM SWITCHES AND CONTROL RELAYS FUNCTION PROPERLY TO OPERATE THE HOPPER PRESSURIZING VALVES.

B. STRUCTURAL INTEGRITY 1. SYSTEM PRESSURE BOUNDRY LEA'< AGE 1ESTING A) SYSTEM PIPING UPSTREAM OF PRESSURIZING VALVES - THE SYSTEM PIPING UPSTREAM OF THE PRESSURIZING VALVES IS CONTINUOUSLY MONITORED FOR LEAKAG" BY THE PRESSURIZING LINE PRESSURE ALARM CIRCUIT. NO NEW SR IS REQUIRED TO VERIFY THE INTEGRITY OF THIS PIPING, WHICH IS NORMALLY PRESSURIZED BY THE HELIUM STORAGE BOTTLES. B) SYSTEM PIPING DOWNSTREAM OF PRESSURIZING VALVES - THE SYSTEM PIPING DOWNSTREAM OF THE PRESSURIZING VALVES IS PRESSURIZED ONCE EACH QUARTER DURING THE SYSTEM TESTS TO VERIFY THAT A CLEAR FLOW PATH EXISTS. NO NEW SR IS REQUIRED TO VERIFY THE INTEGRITY OF THIS Pr'ING, MUCH 0,F WHIC S NOT ACCESSIBLE FOR EXAMINATION. C) SYSTEM PIPING INSIDE REFUELING PENETRATION - ADD A NEW SR WHICH REQUIRES THE SYSTEM PIPING INSIDE THE REFUELING PENETRATION TO BE VISUALLY EXAMINED WHEN THE PENETRATION IT IS CONNECTED TO IS OPENED FOR REFUELING TO VERIFY THE INTEGRITY OF THE SYSTEM PIPING BETWEEN THE PENETRATION CHECK VALVE AND THE CONTROL AND ORIFICE ASSEftBLY H3USING. O s J ) [J J f, f i

PROPOSED SURVEILLANCE REQUIREMENTS REACTOR BUILDING VENTILATION EXHAUST SYSTEM A. OPERATIONAL READINESS 1. SYSTEM TESTING A) LC0 4.5.1 REQUIRES THE REACTOR BUILDING TO BE MAINTAINED SLIGHTLY BELOW ATMOSPHERIC PRESSURE WHENEVER THE REACTOR IS BEING OPERATED OR REFUELED. REACTOR BUILDING PRESSURE IS MONITORED CONTINUOUSLY AND ALARMED IF ABNORMAL TO ENSURE THAT THE REACTOR BUILDING IS CAPABLE OF PERFORMING ITS CONFINEMENT FUNCTION. NO NEW SR IS REQUIRED TO DEMONSTRATE THE CAPABILITY OF THE VENTILATION EXHAUST SYSTEM TO MAINTAIN THE REACTOR BUILDING AT A NEGATIVE PRESSURE. B) CONTINUE EXISTING SR 5.2.21 WHICH REQUIRES THE EXHAUST FAN ACM POWER BUS TO BE ENERGIZED ONCE EVERY SIX MONTHS TO DEMONSTRATE THAT THE ACM FEED BREAKERS AND TRANSFER SWITCHES ARE OPERABLE AND THAT THE EXHAUST FANS ARE CAPABLE OF OPERATION FROM THIS EMERGENCY POWER SOURCE. J s C) ADD A NEW SR WHICH REQUIRES THE NITROGEN STORAGE BOTTLE PRESSURE TO BE MONITORED ONCE EACH WEEK TO ENSURE THAT THE RE0l' IRED ACTUATING [$ PRESSURE IS. MAINTAINED FOR' THE EXHAUST FAN DAMPERS LURING ACM SYSTEM OPERATION. t, 'l s

A.- OPERATIONAL READINESS (CONTINUED) 2. COMPONENT TESTING A) HEPA FILTERS - CONTINUE EXISTING SR 5.5.3E WHICH REQUIRES THE HEPA FILTER PRESSURE DROP TO BE MONITORED ONCE EACH WEEK TO ENSURE THEY ARE CAPABLE OF PASSING DESIGN-AIR FLOW AND ARE NOT EXCESSIVELY PLUGGED. CONTINUE EXISTING SR 5.5.3C WHICH REQUIRES THE HEPA FILTER EFFICIENCY BE D0P TESTED ONCE EACH YEAR OR WHENEVER A FILTER IS REPLACED OR ITS HOUSING REPAIRED TO DEMONSTRATE THAT THE CAPACITY FOR REMOVAL OF PARTICULATES IS ACCEPTABLE. CONTINUE EXISTING SR 5.5.3D.WHICH REQUIRES THE HEPA FILTER BE TESTED WHENEVER THE FILTER HOUSING IS MODIFIED TO DEMONSTRATE THAT THE AIR FLOW IS PROPERLY DISTRIBUTED ACROSS THE FILTER. B) CHARCOAL FILTERS - CONTINUE EXISTING SR 5.5.3E WHICH REQUIRES THE CHARC0AL FILTER PRESSURE DROP TO BE MONITORED ONCE EACH WEEK TO ENSURE THEY ARE CAPABLE OF PASSING DESIGN AIR FLOW AND,ARE NOT EXCESSIVELY PLUGGED. CONTINUE EXISTING SR 5.5.3A WHICH REQUIRES A CHARC0AL SAMPLE TO BE LAB TESTED ONCE EVERY 4400 HOURS (ABOUT SIX MONTHS) TO DEMONSTRATE THAT THE CAPACITY FOR REMOVAL OF RADIOACTIVE METHYL IODIDE IS ACCEPTABLE. CONTINUE EXISTING SR 5.5.3B WHICH i3 REQUIRES THE CHARC0AL FILTERS TO BE TESTED ONCE EACH YEAR OR WHENEVER ] A FILTER BED IS REPLACED OR ITS HOUSING REPAIRED TO DEMONSTRATE THAT THE CAPACITY FOR REMOVAL OF HYDROGENATED HYDROCARBONS IS ACCEPTABLE. e$ CONTINUE SR 5.5.3D WHICH REQUIRES THE CHARCOAL FILTER TO BE TFSTED WHENEVER THE FILTER HOUSING IS MODIFIED TO DEMONSTRATE THAT THE AIR FLOW IS PROPERLY DISTRIBUTED ACROSS THE FILTER BEDS. t>

I S R R O E E T R N P S C K A E O M H A E N H A T E E S A T N D N R W N F O O A T I E M S E H F T A T H N T X O H C S H I T A E U T S I I E H A O S S T S T H E P E C E E X T R Y N R J T E A R I R U I b f. R E U E F U O H E T P Q V Q T H S O E E R E W T N R R E R O E O P E D L T S M H C A H F A E E E C N E O H C R R D H I T H I N T I H G S U O W S R W I N Q T N E I 1 L A S O E I E E M R S 2 T T T 3 D E H S D H T L2 O N B E 5 T C N I I O A5 M A O 5 T H M F N U D W R E R R E S X DS G T S T K R I N O S A C S S A G R N G R E N T I N E H W Y Y I I ) P C E R L T N P T O N E R S U M K S P V E I I E O A E P XC C X B R O EA A E D D E R B ) O D C P E E D E T E A N U H E E U R O S N T H T U N S U E I N S Y D K T G N I I T A S L SE O N N T A T N F E R RT R O I N O R E ES T C S H O C T P P P E S U T C S O M T D E ( U R R A R A E P D E O R E T G 1 H T B T E T A N 1 X L G N A W A R t I A E I N A O R O U T T F F I F T E P T S S G N P C N Y E T N E O T S O F A O L T S I H I S R O M R U D T T U E E E E E T A L C A P V S B D P R N H M L S O N H I E X U O U X A A O O O R N E B F F E D V L T T P O P M ) ) O C D C 2 m sj h n \\

A., OPERATIONAL READINESS (CONTINUED) 3. INSTRUMENTATION AND CONT _ROL TESTING A) EXHAUST FILTERS - ADD A NEW SR WHICH REQUIRES THE EXHAUST FILTER DIFFERENTIAL PRESSURE INDICATION AND ALARM CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE INDICATORS, SWITCHES, AND ALARMS FUNCTION PROPERLY. B) EX11AUST EAtd - ADD A NEW SR WHICH REQUIRES THE EXHAUSY FAN DIFFERENTIAL PRESSURE INDICATION AND ALARM CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE INDICATORS, SWITCHES,.\\ND ALARMS FUNCTION PROPERLY. C) EXHAUST FAN DAMPERS - ADD A NEW SR WHICH REQUIRES THE EXHAUST FAN DAMPER POSITION INDICATION CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE POSITION SWITCHES ACCURATELY INDICATE VALVE POSITION. D) REACTOR BUILDING - CONTINUE EXISTING SR 5.5.1 WHICH REQUIRES THE REACTOR BUILDING PRESSURE INDICATION AND CONTROL CIRCUIT TO BE TESTED i ONCE EACH MONTH AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT 1 THE DIFFERENTIAL PRESSURE TRANSMITTERS, INDICATORS, SWITCH, CONTROLLER AND ALARM FUNCTION PROPERLY. a t N

' 3. IflSTRUMENTATION AIID C0flTROL TESTIflE (CONTINUED) E) EXHAUST STACK - CONTINUE EXISTING SR 5.8.1 WHICH REQUIRES THE EXHAUST STACK RADIATION MONITORING CIRCUIT TO BE TESTED ONCE EACH WEEK, ONCE EACH MONTH, AND CALIBRATED ONCE EACH QUARTER TO DEMONSTRATE THAT THE MONITORS, ALARMS, AND ISOLATION CONTROLS FUNCTION PROPERLY. ADD A NEW SR WHICH REQUIRES THE EXHAUST STACK FLOW INDICATION CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE SENSING ELEMENT AND INDICATORS ACCURATELY MEASURE EXHAUST STACK FLOW. B. STRUCTURAL INTEGRITY - VENTILATION SYSTEMS ARE NOT SUBJECT TO SURVEILLANCE FOR STRUCTURAL INTEGRITY. %me s J 0 '1~ s

PROPOSED SURVEILLANCE REQUIREMENTS REACTOR BUILDING OVERPRESLURE PROTECTION SYSTEM A. OPERATIONAL READINESS 1. SYSTEM TESTING A) CONTINUE EXISTING SR 5.5.2 WHICH REQUIRES THE REACTOR BUILDING PRESSURE RELIEF LOUVERS TO BE EXERCISED ONCE EACH YEAR TO DEMONSTRATE OPERA-BILITY. MODIFY SR 5.5.2 TO REQUIRE THAT THE REACTOR BUILDING PRESSURE RELIEF LOUVERS BE EXERCISED USING THE BACKUP INSTRUMENT AIR HEAD 5R (NITROGEN BOTTLES) ONCE EACH YEAR TO DEMONSTRATE OPERABILITY WITH BACKUP POWER. (NOTE: THIS TEST IS PRESENTLY BEING PERFORMED). 2. COMPONENT TESTING A) PRESSURE RELIEF LOUVERS - MODIFY SR 5.5.2 TO REQUIRE THE PRESSURE RELIEF LOUVERS TO BE TESTED ONCE EACH YEAR IN CONJUNCTION WITH THE SYSTEM TESTS TO DEMONSTRATE THAT THE ACTUATOR STROKES FULLY AND RAPIDLY. 8 Y s J b tO L 4

3. INSTRUMENTATION AND CONTROL TESTING A) REACTOR BUILDING - CONTINUE EXISTING SR 5.5.2 WHICH REQUIRES THE REACTOR BUILDING PRESSURE SENS'fNG LOSIC TO BE TESTED ONCE EACH MONTH AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRES-SURE TRANSMITTERS AND CONTROL LOGIC FUNCTION PROPERLY. B) SUPPLY FAN - MODIFY SR 5.5.2 TO REQUIRE THE SUPPLY FAN INTERLOCK CIRCUIT TO BE TESTED ONCE EACH YEAR IN CONJUNCTION WITH THE SYSTEM TESTS TO DEMONSTRATE THAT THE SUPPLY FAN IS SHUT DOWN WHEN THE LOUVERS ARE ACTUATED. (NOTE: THIS TEST IS PRESENTLY BEING PERFORMED). C) INSTRUMENT AIR HEADERS - ADD A NEW SR WHICH REQUIRES THE PRESSURE INDICATION AND ALARM CIRCUITS FOR THE NORMAL AND BACKUP INSTRUMENT AIR HEADERS TO.BE TESTED AND CALIBRATED ONCE EACH YEAR TC DEMONSTRATE THAT THE PRESSURE INDICATORS, PRESSURE SWITCHES, AND ALARMS FUNCTION

PROPERLY,

] D) LOUVER GROUP HEADERS - ADD A NEW SR WHICH REQUIRES THE PRESSURE ALARM CIRCUITS IN EACH OF THE TWENTY LOUVER GROUP HEADERS TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE SWITCHES AND ALARMS FUNCTION AT THE CORRECT SET PRESSURE. 3 3 -L (In

B. STRUCTURAL INTEGRITY - VENTILATION SYSTEMS ARE NOT SUBJECT TO SURVEILLANCE FOR STRUCTURAL INTEGRITY. -4 8 s J .3 1 1.

PROPOSED SURVEILLANCE REQUIREMENTS PERIFICATION COOLING WATER SYSTEM A. OPERATIONAL READINESS 1. SYSTEMTESTING-CONTINUEEXISTINGSR5.2.21WHICHREQUIRESTHEPURIFkCA-T10tl PUMP ACM POWER BUS TO BE ENERGIZED ONCE EVERY SIX MONTHS TO DEMONSTRATE THAT THE ACM FEED BREAKERS AND TRANSFER SWITCHES ARE OPERABLE AND THAT THE PURIFICATION PUMPS ARE CAPABLE OF OPERATION FROM THIS EMERGENCY POWER SOURCE. x -1

A. OPERATIONAL READINESS (CONTINUED) 2. COMPONENT TESTIfM A) PURIFICATION PUMPS - ADD A NEW SR.WHICH REQUIRES THE PURIFICATION PUMPS TO BE TESTED ONCE EACH QUARTER TO DEMONSTRATE THAT APPLICABLE PUMP PERFORMANCE PARAMETERS ARE ACCEPTABLE. B) COOLING LOOP ISOLATION VALVES - CONTINUE EXISTING SR 5.2'.21 WHICH REQUIRES THE COOLING LOOP ISOLATION VALVES TO BE MANUALLY EXERCISED ONCE EVERY SIX MONTHS TO DEMONSTRATE THAT THE HAND JACK FUNCTIONS PROPERLY. ADD A NEW SR WHICH REQUIRES THE COOLING LOOP ISOLATION VALVES TO BE EXERCISED ONCE EVERY SIX MONTHS TO DEMONSTRATE THAT THE AIR OPERATOR STROKES PROPERLY AND FAILS IN THE PROPER POSITION ON LOSS OF POWER. C) C00 LING' LOOP RELIEF VALVES - ADD A NEW SR WHICH REQUIRES THE COOLING LOOP RELIEF VALVES TO BE TESTED ONCE EVERY FIVE YEARS TO DEMONSTRATE THAT THE VALVES OPEN AT THE CORRECT SET PRESSURE. m J s j J 4.

A'. OPERATIONAL READINESS (CONTINUED) 3. INSTRUMENTATION ANb CONTROL TESTItifi A) PURIFICATION PUMPS - ADD A NEW SR NHICH REQUIRES THE PURIFICATION PUMP DIFFERENTIAL PRESSURE INDICATION CIRCUIT TO BE CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE DIFFERENTIAL PRESSURE INDICATOR AND SWITCH FUNCTIONS PROPERLY. B) ISOLATION VALVES - ADD A NEW SR WHICH REQUIRES THE ISOLATION VALVE POSITION INDICATION CIRCUIT TO BE TESTED ONCE EACH YEAR TO DEMONSTRATE THAT THE POSITION SWITCHES AND INDICATORS FUNCTION PROPERLY. C) PURIFICATION COOLER _ LOOP - ADD A NEW SR WHICH REQUIRES THE PURIFICA-TION COOLER HIGH PRESSURE LOOP ISOLATION CIRCUIT TO BE TESTED ONCE EACH QUARTER TO DEMONSTRATE THAT THE PRESSURE SWITCH FUNCTIONS TO CLOSE THE ISOLATION VALVES AT THE CORRECT SET PRESSURE. B. STRUCTURAL INTEGRITY ,] 1. SYSTEM PRESSURE BOUNDRY LEAKAGE TESTING - ADD A NEW SR WHICH REQUIRES } THE SYSTEM PIPING TO BE EXAMINED FOR LEAKAGE ONCE EVERY YEAR DURING OPERATION TO VERIFY THE INTEGRITY OF THE SYSTEM PRESSURE BOUNDRY. t4 I ~.,

PROPOSED SURVEILLANCE REQUIREMENTS PCRV SAFETY VALVES A. OPERATIONAL READIMESS 1. SYSTEM TESTING - NOT APPLICABLE DUE TO SYSTEM DESIGN 2. COMPONEf1T TESTING A) RUPTURE DISCS - CONTINUE EXISTING SR 5.2.1A WHICH REQUIRES ONE OF THE TWO RUPTURE DISCS TO BE BENCH TESTED EACH YEAR AT SHUTDOWN ON AN, ALTER-NATING BASIS TO DEMONSTRATE THAT THE BELLEVILLE WASHERS DEFLECT AT THE CORRECT SET PRESSURE. B) SAFETY VALVES - CONTINUE EXISTING SR 5.2.1A WHICH REQUIRES ONE OF THE TWO SAFETY VALVES TO BE TESTED IN PLACE EACH YEAR AT SHUTDOWN ON AN ALTERNATE BASIS TO DEMONSTRATE THAT THE VALVE OPENS AT THE CORRECT SET PRESSURE. C) SAFETY ISOLATION VALVES - ADD A NEW SR WHICH REQUIRES EACH ISOLATION ' 'i VALVE TO BE EXERCISED THROUGH A PARTIAL STROKE SEMI-ANNUALLY AND J THROUGH A FULL STROKE ONCE EACH YEAR AT SHUTDOWN TO DEMONSTRATE THAT THE VALVE OPERATOR IS FUNCTIONING CORRECTLY. ADD A NEW SR WHICH RE-2[ QUIRES EACH ISOLATION VALVE TO BE TESTED IN PLACE ONCE EVERY TWO YEARS AT SHUTDOWN TO DEMONSTRATE THAT SEAT LEAKAGE IS ACCEPTABLE. -L ti

A.' OPERATIONAL READINESS (CONTINUED) 3. INSTRUMErlTATI0tl AND C0flTROL TESTING A) SAFETY ISOLATION VALVES - CONTINUE EXISTING SR 5.2.1C2 WHICH REQUIRES THE POSITION INDICATION AND ALARM CIRCUIT TO BE TESTED AND CALIBRATED ONCE EACH YEAR AT SHUTDOWrl TO DEMONSTRATE THAT THE POSITION SWITCHES, INDICATORS, AND ALARMS FUNCTION PROPERLY. B) RUPTURE DISC - CONTINUE EXISTING SR 5.2.1C1 WHICH REQUIRES THE RUPTURE DISC / SAFETY VALVE INTERSPACE PRESSURE ALARM CIRCUIT TO BE TESTED ONCE EACH MONTH AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE SWITCHES AND ALARMS FUNCTION PROPERLY. C) CONTAINMENT TAtlK - ADD A NEW SR WHICH REQUIRES THE CONTAINMENT TANK PREFSURE ALARM CIRCUIT TO BE TESTED AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE PRESSURE SWITCH AND ALARM FUNCTION PROPERLY. D) SAFETY VALVE-ADD A NEW SR WHICH REQUIRES THE SAFETY VALVE BELLOWS RUPTURE ALARM CIRCUIT TO BE TESTED PRIOR TO SAFETY VALVE TESTING TO DEMONSTRATE THAT THE PRESSURE SWITCH AND ALARM FUNCTION PROPERLY. m) "!ATION SHINE MONITOR - CONTINUE SR 5.4.9 WHICH REQUIRES THE DIS-3 RADIATION SHINE MONITOR TO BE TESTED ONCE EACH WEEK AND 3 c.1 u. EACH YEAR TO DEMONSTRATE THAT THE MONITOR FUNCTIONS a or I. u

B.- STRUCTURAL INTEGRITY 1. . SYSTEM PRESSURE BOUNDRY LEAKAGE TESTING A) PCRV PENETRATION PIPING - CONTINUE EXISTING SR 5.2.16 WHICH REQUIRES PCRV PENETRATION INTERSPACE LEAKAGE TO BE MONITORED CONTINUOUSLY AND MEASURED ONCE EACH MONTH TO VERIFY THE INTEGRITY OF THE PCRV SAFETY VALVE PENETRATION PIPING. B) SYSTEM PIPING - ADD A NEW SR WHICH REQUIRES THE NORMALLY UNPRESSURIZED SYSTEM PIPING (BETWEEN THE RUPTURE DISC AND SAFETY VALVE) TO BE EXAMINED FOR LEAKAGE ONCE EVERY TWO YEARS DUP.ING THE SAFETY VALVE SET POINT TEST TO VERIFY THE INTEGRITY OF THIS PIPING. THE INTEGRITY OF NORMALLY PRESSURIZED SYSTEM PIPING IS VERIFIED BY THE CONTAltiMENT TANK PRESSURE ALARM CIRCUIT, WHICH CONTINUOUSLY MONITORS FOR LEAKAGE. 2. STRUCTURE INSPECTIONS A) PCRV SAFETY VALVE TANK - MODIFY EXISTING SR 5.2.1 TO REQUIGE THAT THE TANK CLOSURE FLANGE SEAL LEAKAGE BE DETERMINED ONCE EACH YEAR FOLLOWING TANK CLOSURE TO DEMONSTRATE LEAK TIGHTNESS OF THE CLOSURE SEALS. (NOTE: THIS TEST IS CURRENTLY BEING PERFORMED). MODIFY THIS SR TO ALSO REQUIRE THAT THE TANK CLOSURE BOLTING BE EXAMINED ONCE EACH ' 'i, YEAR TO VERIFY THE ABSENCE OF VISIBLE SURFACE DEFECTS AND THAT THE J TANK WELD JOINTS BE EXAMINED FOR LEAKAGE ONCE EVERY FIVE YEARS DURING THE COVER FLANGE LEAK TEST. => t' f ig 't

PROPOSED SURVEILLANCE REQUIREMENTS PRE-STRESSED CONC. RETE REACTOR VESSEL A. OPERATIONAL READINESS 1. INSTRUMENTS AND CONTROLS A) PENETRATION INTERSPACES - CONTINUE EXISTING SR 5.2.15 AND SR 5.2.16 WHICH REQUIRE THE PCRV PENETRATION INTERSPACE PRESSURE'AND FLOW !N-DICATING CIRCUITS TO BE TESTED ONCE EACH MONTH AND CALIBRATED ONCE EACH YEAR TO DEMONSTRATE THAT THE INSTRUMENTS FUNCTION PROPERLY. B) TENDONS - CONTINUE EXISTING SR 5.2.3 WHICH REQUIRES THE TENDON LOAD CELLS TO BE TESTED AFTER 1, 3, 8, 13, 18, 23, 28 YEARS OF OPERATION TO DEMONSTRATE THAT THE LOAD CELL REFERENCE POINTS HAVE NOT SHIFTED, ADD A NEW SR WHICH REQUIRES THE LOAD CELL ALARM CIRCUIT TO BE TESTED ONCE EACH YEAR TO DEMONSTRATE THAT THE ALARM FUNCTIONS PROPERLY. a 6 % s -j 4

B. STRUCTURAL INTEGRITY 1. PRESSURE BOUNDARY LEAKAGE AND PRESSURE TESTING A) EENETRATION IflTERSPACES - CONTINUE EXISTING SR 5.2.16A WHICH REQUIRES THE PE?lETRATION INTERSPACE LEAK RATE TO BE MEASURED ONCE EACH MONTH TO MONITOR AND VERIFY THE INTEGRITY OF THE PENETRATIONS AND CLOSURES. THE INTERSPACE LEAK RATES ARE ALSO MONITORED CONTINUOUSLY AND ALARMED IF ABNORMAL TO PROVIDE ADDITIONAL ASSURANCE OF INTEGRITY. 2. STRUCTURE INSPECTIONS A) DVERALL STRUCTURE - ADD A NEW SR WHICH REQUIRES THE PRESSURE VS. DEFLECTION CHARACTERISTICS OF THE OVERALL PCRV STRUCTURE TO BE MEASURED AFTER 1, 3, 5 10,15, 20, AND 25 YEARS OF OPERATION TO VERIFY THAT THE RESPONSE IS ELASTIC AND THAT NO SIGNIFICANT PERMANENT STRAINS EXIST. I s J D D (9 s

- 2. STRUCTURE INSPECTIOHS (CONTINUED) B) CONCRETE SURFACES - CONTINUE EXISTING SR 5.2.4 WHICH REQUIRES THE TOP, BOTTOM, AND SIDE SURFACES OF THE PCRV CONCRETE TO BE VISUALLY EXAMINED AFTER 1, 3, 13, AND 23 YEARS OF OPERATION TO VERIFY STRUCTURAL INTE-GRITY BY ASSESSING THE SURFACE CRACKING PATTERNS AND SIZES. C) CONCRETE HEADS AND WALL - CONTINUE EXISTING SR 5.2.13 WHICH REQUIRES THE HELIUM PERMEATION RATE OF THE CONCRETE IN THE PCRV HEADS AND SIDE WALL TO BE MEASURED AFTER 3, 8, 13, 18, 23, AND 28nYEARS OF-OPERATION, AND THAT ANY CHANGES IN THIS RATE BE ASSESSED TO VERIFY STRUCTURAL INTEGRITY. D) PCRV LINER - CONTINUE EXISTING SR 5.2.5 WHICH REQUIRES THE CHARPY IMPACT SPECIMENS, PLACED ADJACENT TO THE OUTSIDE SURFACE OF THE TOP HEAD LINER, TO BE TESTED AFTER 5, 15, AND 25 YEARS OF OPERATION TO VERIFY THAT CHANGES IN NOTCH TOUGHNESS DUE TO IRRADIATION ARE ACCEPTABLE. CONTINUE EXISTING SR 5.2.14 WHICH REQUIRES SELECTED AREAS OF THE PCRV LINER WALL THICKNESS TO BE MEASURED BY ULTRASONIC TESTING AFTER 3, 5, 15, AND 25 YEARS OF OPERATION TO VERIFY THAT 3 CHANGES IN WALL THICKNESS DUE TO CORROSION ARE ACCEPTABLE. s J w O

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2. STRUCTURE INSPECTIONS (CONTINUED) E) PCRV TENDON ANCHORS - ADD A NEW SR WHICH REQUIRES 5% OF THE PCRV TENDON END ANCHOR ASSEMBLIES TO BE VISUALLY EXAMINED AFTER 1, 3, 8, 13, 18, 23, AND 28 YEARS OF OPERATION TO VERIFY THAT NO CORROSION DAMAGE OR OTHER EVIDENCE OF CHANGES IN STRUCTURAL INTE-GRITY EXISTS. F) PCRV TENDON WIRES - CONTINUE EXISTING SR 5.2.2 WHICH REQUIRES 1% OF THE TENDON TUBE WIRE SAMPLES TO BE VISUALLY EXAMINED AFTER 1, 3, 8, 13, 18, 23, AND 28 YEARS OF OPERATION TO VERIFY THAT NO CORROSIOM DAMAGE EXISTS. G) PCRV TENDON TUBES - CONTINUE EXISTING SR 5.2.2 WHICH REQUIRES A SAMPLE OF THE ATMOSPHERE IN A REPRESENTATIVE NUMBER OF TENDON TUBES (WITH AND WITHOUT LOAD CELLS) BE TAKEN AND ANALYZED AFTER 1, 3, 8, 13, 18, 23, AND 28 YEARS OF OPERATION TO VERIFY THAT CORROSION PRO-DUCTS DO NOT EXIST. 'i H) TENDON LOADS - SELECTED TENDON PRESTRESS FORCES ARE MONITORED 1 CONTINUOUSLY AND ALARMED IF ABNORMAL TO ENSURE THAT THE REQUIRED LOADS ARE MAINTAINED AND THAT AN OVERLOAD CONDITION DOES NOT EXIST. [}$ NO NEW SR IS REQUIRED TO ASSURE THAT TENDON LOADS ARE ACCEPTABLE. i

2. STRUCTURE IflSPECTIONS (CONTINUED) I) REFUELING PENETRATIOff LIMIT STOPS - ADD A NEW SR WHICH REQUIRES THE REFUELING PENETRATIONS HOLD DOWN PLATE BOLTING TO BE VISUALLY EXAMINED AT EACH REFUELING SHUTDOWN TO VERIFY THE INTEGRITY OF THE BOLTING. J am & 4 e h I,

PROPOSED SURVEILLANCE REQUIREMENTS HELIUM CIRCULATOR SHUTOFF VALVES ADD A NEW SR WHICH REQUIRES EACH OF THE FOUR HELIUM CIRCULATOR ~ SHUT 0FF VALVES TO BE MONITORED ONCE EACH YEAR TO ENSURE THAT THE VALVE CLOSES PROPERLY WHEN ITS ASSOCIATED CIRCULATOR IS SHUTDOWN. HOIE EXISTING GilIDELINES PROVIDED IN THE TECHNICAL SPECIFICATIONS FOR REPORTING UNUSUAL EVENTS ARE CONSIDERED ADEQUATE TO ENSURE THAT ANY OBSERVED FAILURE OF A HELIUM CIRCULATOR SHUT 0FF VALVE TO CLOSE PROPERLY WILL BE REPORTED TO THE NRC. 0 x J +wwb fn

Gy MEETING AGENDA HELIUM CIRCULATOR AUXILIARY SYSTEM MODIFICATIONS I. INTRODUCTION A, BACKGROUND B. SYSTEMS AFFECTED II. DESIGN OBJECTIVFa

III, PROPOSED MODIFICATIONJ.

A. JUSTIFICATION e 17 17 J e

53 l BUFFER HELIUM UPSET PREVENTION II. FOUR DESIGN OBJECTIVES WERE DEFINED: 1 1) MINIMIZE THE POSSIBILITY OF WATER ENTERING THE BUFFER HELIUM PATH. 2) IMPROVE CONTROL OF LOW PRESSURE SEPARATOR WATER LEVEL UNDER UPSET CONDITIONS. 3) MINIMIZE THE POSSIBILITY OF ABNORMAL HIGH HELIUM PRESSURE IN THE LOW PRESSURE SEPARATOR AND AT THE HELIUM RECOVERY COMPRESSOR SUCTIONS. 4) MINIMIZE AND CONTROL THE RELEASE OF RADIOACTIVE GAS AND WATER INTO THE REACTOR BUILDING. /

59 1) MINIMIZE THE POSSIBII.ITY OF WATER ENTERING THE BUFFER HELIUM PATH MODIFICATION POSITIVE FEATURES 1.1 PELIUM RECOVER COMP. K.0. POT INCREASE PRESENT DRAIN LINE AND INCREASES LINE CAPACITY TO HANDLE CONTROL VALVE FROM K.0, POT TO MAXIMUM POSTULATED WATER LEAK. BEARING WATER SURGE IANK TO l-1/2". MAKE K.0, POT LEVEL CONTROL VALVE PROVIDES POSITIVE ISOLATION BETWEEN (LV 21385) SHUT AUTOMATICALLY WHEN THE BEARING WATER SURGE TANK AND THE VENT VALVE HV-21334-1 IS OPEN. REACTOR BUILDING VENTILATION EXHAUST DUCTING. ADD A VALVED MANUAL DRAIN AND A PROVIDES'A POSITIVE LOCAL INDICATION K.0. POT GAGE GLASS. AND MEANS OF CONTROL FOR K.0, POT WATER LEVEL, AS WELL AS PROVIDING A BACKUP TO AUTOMATIC VALVE. ADD A HIGH LEVEL ALARM WHICH IN-GIVES CONTROL ROOM INDICATION WHEN DICATES LOCALLY, AND IN THE WATER LEVEL IS HIGH. CONTROL ROOM. 9/ t ^

sc 1) MINIMIZE THE POSSIBILITY OF WATER ENTERING THE BUFFER HELIUM PATH MODIFICATION POSITIVE FEATURES 1.2 HELIUM DRYER K.O. POT un wu 7-ADD A NEW AUTOMATIC LEVEL CONTROL PROVIDES AUTOMATIC DRAIN IN -PLACE-OF-VALVE TO K.0. POT DRAIN LINE. A MANUAL DRAIN WHICH REQUIRED SPECIFIC OPERATOR ACTION. THIS LINE IS ALSO SIZED FOR THE MAXIMUM POSTULATED WATER LEAK. RETAIN THE EXISTING K.0. POT MANUAL PROVIDES A POSITIVE LOCAL INDICATION DRAIN AND GAGE GLASS PROVISION. AND MEANS OF CONTROL FOR K 0. POT WATER LEVEL, AS WELL AS PROVIDING A BACKUP TO AUTOMATIC VALVE. ADD A HIGH LEVEL ALARM WHICH IN-GIVES CONTROL ROOM INDICATION WHEN DICATES LOCALLY AND IN THE CONTROL WATER LEVEL IS HIGH. ROOM. 1.3 BUFFER HELIUM RECIRCULATOR IANK FLOODING AND BLOWDOWN PROVIDE AUTOMATIC SHUTDOWN AND ISOLATES THE RECIRCULATOR EXPERID4CING ISOLATION OF A BUFFER HELIUM A HIGH WATER LEVEL PROBLEM FROM THE RECIRCULATOR COINCIDENT WITH A REST OF THE SYSTEM. WATER HIGH LEVEL ALARM IN THE RECIRCULATOR CONTAINMENT TANK. PROVIDE A CHECK VALVE IN THE THESE CHECK VALVES WILL PREVENT A DISCHARGE LINE FROM EACH SHORT-CIRCUIT FLOW PATH THROUGH THE ~ BUFFER HELIUM RECIRCULATOR. BEARING WATER SURGE TANK IF A RECIRCUL4TC CONTAINMENT TANK LEVEL CONTROL VALVE FAILED OPEN jq., .)f

.I 2) IMPROVE CONTROL OF LOW PRESSURE SEPARATOR WATER LEVEL.UNDER UPSET CONDITIONS MODIFICATION POSITIVE FEATURES 2.1 LOW PRESSURE SEPARATOR DRAINS DECOUPLE THE POSITIONING OF THE ENSURES A DRAIN PATH FROM THE LOW SYSTEM 62 (RADI0 ACTIVE LIQUID PRESSURE SEPARATOR TO SYSTEM 62 WASTE) DRAIN VALVE RV-21251-1 DuRING A HIGH RADIATION CONDIT0N, FROM HS-21252. INCREASE EXISTING PIPING AND PROVIDES ADEQUATE CAPACITY FOR THE ISOLATION VALVE IN THE RADI0 ACTIVE MOST SEVERE CONDITION, LIQUID WASTE DRAIN PATH TO 3" SIZE MATERIAL, b

L 3)

MINIMIZE THE POSSIBILITY OF ABNORMAL HIGH HELIUM PRESSURE IN THE LOW PRESSURE SEPARATOR AND AT THE HELIUM REC 0VERY COMPRESSOR SUCTIONS. MODIFICATION POSITIVE FEATURES 3.1 ISOLATION OF HELIUM PURGE RETURN FLOW FROM DRYER ADD A NEW SHUT 0FF VALVE IN THE AUTOMATICALLY ISOLATES A 10 ACFM HELIUM PURGE LINE LEAVING THE FLOW TO THE LOW PRESSURE SEPARATOR DRYER. THIS VALVE SHALL BE AUTO-WHEN THE PRESSURE IS ABOVE ITS NORMAL MATICALLY SHUT WHEN LOW PRESSURE LEVEL OR WHEN RADIATION IS DETECTED. SEPARATOR PRESSURE INCREASES TO 35 PSIG, OR WHEN RADIATION IS DETECTED IN THE LOW PRESSURE SEPARATOR. ADD A LOCAL INDICATOR LIGHT TO PROVIDES THE OPERATOR WITH LOCAL SHOW WHEN THE NEW VALVE IS SHUT. INDICATION THAT THE VALVE IS CLOSED.

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G3 4) MINIMIZE AND CONTROL THE R._ LEASE OF RADI0 ACTIVE GAS AND WATER INTO THE REACTOR BUILDING MODIFICATION POSITIVE FEATURES 4.1 RELIEF VALVE DISCHARGE EXTEND THE TAILPIPES OF THIS PREVENTS THE RELEASE OF RELIEF VALVES V-211266 AND POTENTIALLY RADIOACTIVE HELIUM V-21345 TO THE REACTOR BUILDING DIRECTLY INTO THE REACTOR BUILDING. VENTILATION EXHAUST DUCTWORK. INSTEAD, IT IS DIRECTED TO THE REACTOR BUILDING EXHAUST FILTERS. ACCESS TO THE REACTOR BUILDING WILL BE MAINTAINED. 4.2 RELIF VALVE SETTINGS CHANGE THE SETTING OF RELIEF THIS PROVIDES A STAGED RELIEF. VALVE V-211266 TO 40 PSIG. CHANGE THE SETTING OF RELIEF VALVE V-21478 TO 45 PSIG. RELIEF SETTING OF V-21345 TO REMAIN AT 50 PSIG. t 7 ' ; o

4) MINIMIZE AND CONTROL THE RELEASE OF RADI0 ACTIVE GAS AND WATER INTO THE REACTOR BUILDING MODIFICATION POSITIVE FEATURES 4.3 PROHIBIT OPENING HV-21334-1 DuRING A HIGH RADIATION CONDITION PROVIDE AN ELECTRICAL SIGNAL PREVENTS OPERATOR ACTION FROM FROM RT-21251 WHICH PROHIBITS OPENING THE VALVE DURING A RADIATION HV-21334-1 FROM OPENING DURING INCIDENT. A RADIATION INCIDENT. THIS OVER-RIDES HS-21334. ADD A VALVE POSITION INDICATOR CONTINUOUSLY ADVISES THE OPERATOR IN THE CONTROL ROOM FOR OF THE VALVE POSITION. HV-21334-1. 11 q O

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-#h.:,.dh',} 4C;[(g, '8 ' ""EA" ..u-.; 1,i (NE-1133) hf_it3gg LOCDEftCTOR k OETICT0ft LOCATION ISV core layou:. and locations of ex-core detectors a + P00R 018iRll

63 FSV INITIAL CORE: DETECTOR RESPONSE TO SEQUENTIAL ROD BANK WITilDRAWAL i t IChannelIndicatedPower h(a) Av Source True Power ) of og Bank (s) f Withdrawn III IV V VI VII VIII Six Data ~3A 1.45 1.38 1.19 0.90 0.87 0.96 1.13 Calculation 1.3 1.3 1.1 0.9 0.9 1.0 1.08 Measurement 4D 0.98 0.97 0.96 1.65 1.80 1.77 1.36 Calculation 0.98 0.98 0.97 1.54 1.65 1.64 1.29 Measurement 3B 0.88 0.81 0.86 0.99 1.07 0.93-0.92 Calculation 0.94 0.89 0.93 0.97 1.02 0.99' O.96 Measurement 4E 1.13 1.19 1.00 1.01 1.17 1.30 1.13 Calculation 1.18 1.12 1.02 1.13 1.22 1.35 1.17 Measurement 4A 1.78 1.60 1.57 0.93 0.91 0.96 1.29 Calculation -4C 0.93 0.93 0.92 1.58 1.43 1.36 1.19 Calculation 3D. 0.92 0.89 1.02 0.91 0.93 0.85 0.92 Calculation Reg rod 0.97 0.97 0.98 0.98 0.97 0.97 0.97 Calculation -(115-190 in.) .3h+4D 1.27 1.27 1.07 1.39 1.48 1.64 1.35 Measurement 4D+3B 0.92 0.87 0.90 1.49 1.68 1.62 1.25 Measurement 3B+4E 1.11 1.00 0.95 1.10 1.24 1.34 1.12 Measurement) Meas / calc (b 4 E-4 A 2.10 1.79 1.60 1.05 1.11 1.30 1.49 4 A+4 C 1.65 1.49 1.44 1.47 1.30 1.31 1.44 Calculation 40+3D 0.86 0.83 0.94 1.44 1.33 1.16 1.09 Calculation 3n+RR 0.89 0.86 1.00 0.89 0.90 0.32 0.89 Calculation 3A+4 D+3B 1.19 1.13 1.00 1.35 1.51 1.62 1.30 Measurement 4D+3B+4E 1.09 0.98 0.92 1.69

2. ' s 2.19 1.49 Measurement Meas / calc ((b) 3B+4 E+4 A 1.97 1.59 1.49 1.02 1.13 1.28 1.41 Meas /cale b) 4E+4A+4C 1.93 1.67 1.47 1.66 1.59 1.76 1.68 4A+4C+3D 1.52 1.32 1.47 1.34 1.21 1.11 1.31 Calculation 4C+3D+RR 0.83 0.80 0.92 1.41 1.29 1.12 1.06 Calculation r-s It is assumed that all detectors are calibrated to read'hcat-balance power when the bank (s) are fully inserted. The figures represent the decalibration that would occur when the bank (s) are fully withdrawn if there were no intermediate recalibration of the detectors.

( Mcas/cale indicates that measured decalibration factors were used wherc available, and calculational results were used where no measurements e were available. /

ta 1. REACTOR WILL TRIP AT OR BELOW l@% ACTUAL REACTOR POWER. 2. OPERATOR HAS AN ACCUPATE INDICATION OF ACTUAL REACTOR POWER LEVEL. 3. REACTOR CONTROL NOT AFFECTED BY INACCURACIES IN REACTOR PO!'ER fEASUREMENT. C0flCEf0S 3o 7 ',

il 1. ADDITIONAL CALIBRAITON OF PPS LINEAR NUCLEAR OiANNELS NEEDED AS A FUNCTION OF LOAD. 2. MUST PROVICE OPERATOR WITH CORRECTED REACTOR POVER INDICATION. 3. MUST REDUCE REACTOR TRIP SET POINT TO ASSURE REACTOR TRIP AT 5 ll10% REACTOR POWER, 11. NO EFFECT ON PLANT CONTROL. RESULTS OF EVALUATI@ i]/

li., 1. WITH Tm ISS IN THE STARTUP NDE, A CALIBRATION IS TO BE DONE WHEN HEAT-BALANCE POVER IS BETWEEN 2% AND 4% OF RATED POVER. 2. WHEN INCREASING POVER WITH THE ISS IN THE LGv P0wEa MODE, A CALIBRATION IS TO BE DONE WHEN HEAT-BALN4CE POWER IS BETWEEN 24% AND 28% OF RATED POWER. 3. WHEN DECREASING POWER WITH THE ISS IN THE PGVER WDE, A CALIBRATION IS TO BE DONE WHEN HEAT-BALANCE POWER DROPS BELOW 36% OF RATED PGiER. ADDITIOVL CAllBPATION REQUIRBE'TS ./"

75 UNCORRECTED t FLUX AVERAGEPOWER FLUX MW(T) HR CONTROLLER INTEGRATOR METER (" (" I (NC11SS) PREVIOUS { 103 103 103 e i t HEAT BA1.ANCE DIGITAL TO. FLUX RECORDER ANALOG CALCULA-4 (NR11SS) (ALSO T'JN (DATA CONVERTER RECORDS P/F) LDGGER) (DATA LOGGER) ^ 103 CDP.RECTED INPUT ONLY ON AVE". AGE F0WER COMPUTER FAILURE V V DIV! DER / POWER / FLOW AMPLIFIER MEASUREMENT CIRCUlT (XMS 11282) 103 110 l ~ l CORRECTED AD0ITIONAL HEAT POWER BALANCE CIRCUITRY METER IN BOX Heat balahce calibration jg ..,g

74 130 REACTOR TRIP MAY OCCUR AT TRUE POWER EXCEEDING 140% 120 110 100 90 g $2 80 E FE -70 2b REACTOR TRIP ALWAYS OCCURS 00 g ATTRUE POWER < 140% E 50 40 30 20 - 10 I I I I I I I 0 O 10 20 30 40 50 60 70 80 SO 100 INDICATED POWER (%) t

-73 120 110 100 SD 80 2m TRIP 70 n.*p RVIP E 60 2ti 50 40 30 20 10 I I' I I I I I I t i 0 10 20 30 40 50 00 70 80 SD 100 INDICATED POVIER (%) ~ '~ Recommended program for trip and RWP setpoints ?4 3 <3 ?

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LOW HOLD CIRCUIT IFIER MODULE SETPOINT SUMMER l SETPOINT l l A A l 15-60% 0 - 30% l l OFFSET OFFSET I I I-REACTOR TRIP SETPOINT l -1 L--- I g ROD WITH0RAWAL PROHIBITSETP0lfR V I L_ y PROGRAMMABLE HIGH NUCLEAR. _l j BISTABLE TRIP POWER TalP ADDITIONAL FLOATING TRIP PalNT CIRCUITRY y IN BOX g PROGRAMMABl.E ROD WITHDRAWAL y y s BISTABLE TRIP PROHIBIT J A PPS floating trip circuit (typical of six places) J l %l Q-

I t+- 1% i t t l o, re I 140 r_ TRIP S8TP0lNT 'e i I ne 120 I SETPOINT HOLDS DUE TO HIGH RATE OF INCREASE jeg 0F MEASURED F0WER AND U-TRIP OCCURS AT 70% F0WER E 5:: og 80. - Oy + a S 60 LE U W o 5 INDICATED ROD WITHDRAWAL. ' REACTOR ACCIDENT FR0M 40 20% POWER y F0WER 7 20 0 TIME FSV floating reactor trip on high reactor power (example of operation showing startup, load-changing, load reduction to 20% power, and rod withdrawal ac'cident) e 9

E6 ~ PHYSICAL MODIFICATIONS 1. ELECTRONICS MODIFICATION TO PPS LINEAR POWER CHANNELS TO INCORPORATE SLIDING REACTOR IRIP AND R0D WITH DRAWL PROHIBIT SETPOINTS. 2. ELECTRONICS MODIFICATION TO PROVIDE OPERATORS WITH CORRECTED REACTOR POWER INDICATION. ADMINISTRATIVE 1. IMPLEMENT ADDITIONAL CALIBRATION REQUIREMENTS AS A FUNCTION OF REACTOR POWER.

SUMMARY

OF PROPOSED RESOLUTION 4

79 e PCRV LINER COOLING SYSTEM "A,m,. 1 ) !- h '. w HEAT LOAD (X 106 BTU /HR) 60% POWER DESIGN ALLOWABLE PCRV SUBSECTION LOOP 1 LOOP 2 AT 100% Top Penetrations .618 .484 2.270 Core Support Floor Top 1.29 1.25 Core Support Floor Side .774 1.04 Core Support Floor Ecttom .296 .357 Total Core Suppcrt Floor 2.360 2.647 8.076 Upper Barrel and Top Head 1.17 1.28 6.292 Lower Barrel 1.07 1.18 ~3.350 Bottom Head and Bottom .Pene tra tio ns 1.46 1.38 4.887 Total by Loon 6.678 6.971 Total PCRV 13.649 24.87 l / } / */ "9 [' e

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~. 6 STU/HR) HEAT LOAD (X 10 70% POWER DESIGN ALLOWABLE PCRV SUBSECTION LOOP 1 LOOP 2 AT 100% Top Penetrations 1.01 0.86 2.270 Core Support Floor Top 1.37 1.15 Core Supnort Floor Side .86 .96 Core Suoport Floor Bottom .31 .32 Total Core Support Floor 2.54 2.43 '8.076 Upper Barrel and Top Head 1.41 1.30 6.292 Lower narrel 1.11 1.02 3.350 Bottom Ucad and Bottom Penetrations 1.60 1.43 4.887 Tota] by Loop 7.67 7.04 12.435 _ Total PCRV 14.71 24.87

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e' Table 1-1 PCRV Liner / Concrete Hot Spots Local Maximum Concrete Temperature Effective Concrete at the Temperature at the Liner Cooling Liner / Concrete Liner / Concrete Tube aT Interface

  • F Interface
  • F p.e /...A W3, h+

l h

1) Top Head 20*F 0 55%

20l*F 0 100% 128*F 0100% (

f Penetrations 30^F 0 100%
  1. 4
  • F @ ( 6 7.

on 8-/S-79 n oA.

2) Core Outlet 20*F 0 80%

200*F'O 20% ' 120*F O 100% o IC - Thermocouple 21*F 0 100% 346*F 0 100% Penetrations er : 2.3

  • F (d' %

o*. S 19 - ( Ha3 h**h 6 I .f% <ei.u u. 4. u ..w, s -r o, )c,

3) Core Barrel Seal 40*F060%(I) 250*F 0 27%

120*F 0 100% )

  • r @ i t %
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IT-7f 3 100% Area 47*F 0 10 n*rto 4) Peripheral Seal 20*F 0 40% 200*F 0 100% 120*F 0100% Area

  • F O lo g_yg_yg

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5) Loop Divider 15*F 0 100%

200*F 0 20% 120*F 0 100% Baffle Area . c,, 4 5. g,,,,,,,j, 354*F 0 100%

6) Steam Generator 16*F,0 100%

200*F 0 40% 120*F 0 100% Penetrations 220*F 0 100%- O ' I ' 3.03*Fg 907, on 8-/7-7c1 7) Cross Over 3*F 0 100% 200*F 0 11% 110*F 0 100% Pipe 300*F 0 100% J II) 150 Normal 130'F FSAR Stated 20*F 200 Normal Hot Spot (I) 4 Value =, (1) Amendment 29 established 40*F cooling tube AT, 250*F local maximum concrete temperature at the liner / concrete interface for the core support floor. Amendment 29 also allows under certain ci rri'n t nre - n incal m wimeim ennerata t-ar d 'ra a f "a *F

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