ML19257B288

Forwards Response to NUREG-0578.Status of Items & Repts Re Inadequate Core Cooling Guideline Development Encl

ML19257B288
Person / Time
Site:	Rancho Seco
Issue date:	01/07/1980
From:	Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Reid R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8001150496
Download: ML19257B288 (84)

Text

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) SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S stree SMUD 95813; (916) 452 3211 Jz. u.

7, 1979 Mr.Ro M n. Reia Operating Reactor Branch No. 4 Division of Operating Reactor U. S. Nuclear Regulatory Commission Washington, D c.

20555

Dear Mr. Reid:

Tne Sacramento Municipal Utility District herein provices status on actions required by NUREG-0578. Attachmant I liets the status of each item.

Attachrnents 2, 3 and 4 are related to Inadequate Core Cooling Guideline development as discussed in response to item 2.1.3.b of NUREG-0578. The District requests prompt review and approval of these attachments so that guidelines can ba incorporated into procedure s.

Sincere ly, s

5

1. Lti k e John Mattimoe Assistant General Manger and Chief Engineer 1751 001 ll Del 80011504 94 gg

[

• 8 AN ELECTRIC SYSTEM SERV!NG MORE THAN 600,000 IN THE HEART OF C A LIFO RNI A

2.1.1 Emergency Power Supply - Pressurizer Heaters See the District's November 26, 1979 letter to you.

2.1.2 Emergency Power Supply - PORV 81ock Valve, PORV and Pressurizer Level See the District's letter to Mr. D. G. Eisent.ut of November 26, 1979.

2.1.3 Performance testing for BWR and PWR Relief and Safety Valves.

By letter dated December 17, 1979, Mr. William J. Cahill, Jr.,

Chaiman of tne EPRI Safety and Analysis Task Force submitted

" Program Plan for the Perfomance Verification of PWR Safety / Relief Valves and Systems".

The Sacramento Municipal Utility District considers the program to be responsive to the requirements presented in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Tem Recommendations" dated July,1979, Item 2.1.2 which recommended in part, " commit to provide perf omance verification by f ull scale prototypical testing for all relief and safety valves.

Test conditions shall include two-phase slug flow and subcooled liquid flow calculated to occur for design basis transients and accidents."

The EPRI Prcgram Plan p ovides for completion of the essential portions of tne test program oy July,1981. The District will be participating in the EPRI program to provide program review and to supply plant specific data as required.

2.L 3.a Direct Indication of Power - Operated Relief Valve and Safety Valve Position Indication for PWR's and BWRs.

As described in the Districts letter of November 26, 1979 an acoustic monitoring system will be installed for the PORV and the safety valves. We have decided not to make the systen a safety grade sy stem.

It will be powered from a vital instrument bus.

It will be seismically qualified in a manner consistent with the system to which it is attached.

Procedures will be available for backup indication.

2.1.3.0 Instrumentation for Detection of Inadequate Core Cooling in PWR's and 8WR's.

The District has received all input from Babcock and Wilcox (B&W) concerning inadequate core cooling guidelines. The first set of guidelines were sent to you in the Districts November 26, 1979 letter. The guideline cases in the November 26, 1979 submittal included loss of RCS inventory without reactor coolant pumps operating, loss of RCS inventory with reactor coolant pumps operating, and loss of natural circulation due to loss of heat sink.

These three cases were incorporated into the guidelines titled "Small Break Operating Guidelines for Oconee 1, 2, and 3; Three Mile Island I and 2; Crystal River 3; and Rancho Seco 1 (B&W Document No.

1751 002 s

69-1106001-00 November 1979)" and submitted as Attachment 2 to the r

District's November 26, 1979 letter.

In the same letter we submitted

" Analysis Summary in Support of Inadequate Core Cooling Guidelines for a Loss of RCS Inventory (8&W Document No. 86-1105508-00) as.

The District covers the loss of decay heat removal case by submittal of Attachment 2 to this letter. Attachment 2 is titled " Inadequate Core Cooling Decay Heat Removal System Mode of Operation (BW Document No. 69-1106921-00 December 1979). Attachment 3 to this letter titled " Analysis Summary in Support of Inadequate Core Cooling Guidelines (B&W Document No. 86-1105508-01)" provides all of the analysis information contained within the analysis summary sent as attachment 3 to the District's November 26, 1979 letter, and it also contains the analysis summary of the loss of decay heat removal case.

At September 13, 1979 meeting with the BW Owner Group, the NRC staf f requested guidelines for departure from nucleate boiling (DNB) at power. The resulting analysis shows that DNB at power is not a credible situation. Tnerefore, BW provided for operators an explanation of why DNB at power is not credible.

The explanation to the operator and the supporting analysis summary are included as to this letter.

Historically the NRC staff has approved changes to the small break operating guidelines. The District will not incorporate the guidelines into existing operating procedures until the NRC staff approves the guidelines. NRC staff review and approval is requested.

Subcooling Meter The District will be unable to obtain safety grade wide range temperature indication in the near term.

In addition, in the near tenn we will not provide separate power supplies to the subcooling meters. We will upgrace subcooling meters to safety grade during the refueling outage of 1981.

This inf ormation modifies the District's response of November 26, 1979.

2.1.3.c Instrumentation for Detection of Inadequate Core Cooling - Additional Instrumentation.

See the District's November 26, 1979 letter to you.

2.1.4 Containment Isolation Provision for PWR's and BWR's.

Rancho Seco containment isolation systems currently isolate on diverse parameters -- low reactor coolant system pressure (1600 psig) and hign containment pressure (4 psig).

In addition, reset of either of the containment isolation signals does not result in automatic reopening of containment isolation valves. As such, modification of the system wit.a respect to isolation initiation or reset action is not necessary. The systen controlling the isolation is the Safety Features Acutation System (SFAS).

1751 003 The District perf omed a detailed review of systems associated with each containment penetration. The review defined essential and nonessential systems.

Essential systems are those systems required immediately after an SFAS initation such as high pressure injection or they are systems whose continued operation will not cause accident recovery problems and whose continued operation may aid in accident recovery such as the reactor coolant pump seal supply lines. Essential systems are open across the reactor building boundary af ter an SFAS initiation. Nonessential systems are those systems not required immediately af ter the SFAS initiation sucii as the reactor building pressure equalization line or the reactor building emergency sump suction lines. Time for operator control of nonessential systems is available.

These systens are either isolated or remain isolated af ter an SFAS initiation.

A system by system evaluation f ollows:

DECAY HEAT REMOVAL SYSTEM The Decay Heat Removal System serves to remove decay heat from the reactor core during low pressure reactor coolant system operation..The low pressure situations may be either normal or emergency operations.

In the emergency situation, the system injects water into the reactor vessel. The need for this low pressure injection is immediate. The system is essential and the low pressure injection valves open af ter an SFAS signal.

f uring the low pressure injection mode of operation, the system takes a suc*. ion on tne borated water storage tank (BWST) until the tank reaches its low-low alarm setpoint. Only af ter the BWST reaches this level is suction shif ted to the reactor building errergency sump. The shif t is perf omed by the plant operator from the control room. Therefore, the emergency sump suction valves are in a nonessential portion of the sy st em. They are closed during nomal operation.

In the normal decay heat removal mode, the system takes a suction on the reactor coolant system. The suction section of this system for nomal decay heat removal is nonessen"'al.

Locked closed valves isolate the reactor building penetration.

NUCLEAR SERVICE COOLING WATER SYSTEM The nuclear service cooling water system is a closed loop system serving to transfer heat from the Decay Heat Removal Coolers and Reactor Building Emergency Coolers to the Nuclear Service Raw Water System. Decay Heat Removal Coolers are outside of the reactor building.

The Nuclear Service Cooling Water System water lines which penetrate the containment provide cooling for the Reactor Building Emergency Coolers.

The penetration valves must be open immediately af ter a loss of coolant accident (LOCA) to afford proper post-accident containment atmosphere cooling. Therefore, the system is an essential system. The valves are currently opened by an SFAS signal.

1751 004 REACTOR BUILDING SPRAY SYSTEM Upon occurrence of a LOCA, the Reactor Building Spray System sprays borated water into the Reactor Building atmosphere. The spray cools the atmosphere and reduces the post-accident temperature and pressure within the builoing. Simultaneously the spray additive reduces, by enemical reaction, the post-accident level of fission products in the Reactor Building atmosphere. Accident analyses assume operation of the system within 5 minutes of a LOCA so an operator may not have time to manually initiate the system. Thus, the system is essential for accident mitigation and it is classified as an essential system. The system's motor operated valves are opened by an SFAS signal.

MAKEUP AND PURIFICATION SYSTEM The Makeup and Purification System is conveniently divided into four segments for essential / nonessential analysis.

The first section is the letdown portion of the system. This portion of the system is not used in accident situations and therefore it is not an essential portion of the system.

It is isolated by an SFAS signal.

The second section is the high pressure injection portion of the sy stem. Hign pressure injection supplies water for Reactor Coolant System inventory and pressure control during accident situations. This section of the Makeup and Purification System is tnerefore an essential section.

System valves open on an SFAS signal.

The third section is the reactor coolant pump (RCP) seal supply section of the system. RCPs and RCP seals are not necessary for recovering from any accident situation. Never-the-less, there will be situations in which RCP operation may be desirable af ter an SFAS actuation. Long tenn operation of the pumps without seal flow is not advisable.

The RCP seal supply section does satisfy 10CFR50, Appendix A, Criterion 55 (Reactor Coolant Pressure Boundary Penetrating Containment).

The last section is the RCP seal return section. RCP's may be operated without seal return flow. Therefore, this section of the system is nonessential and it is isolated by an SFAS actuation.

FEEDWATER SYSTEMS Feedwater systems are conveniently divided into two categories.

The first system segment is the main feedwater system. Operation of this segment of the system is not assumed for accident recovery.

Therefore, the system is a nonessential system.

The system nas no connection with the Reactor Coolant System or witn other systems containing radioactive fluids. Main feedwater system boundaries with systems.ontaining radioactive fluids are routinely monitored for integrity.

The main feedwater system is not isolated on an SFAS signal.

1751 005 The second r: ' a segment is the auxiliary feedwater system. The small break anaY ussumes operation of the systen. Thus the auxiliary feedwater porticn of the system is essential.

The system flowpath is opened by an SFAS signal.

WATER GAS SYSTEM The Waste Gas Systen has no direct penetration into the reactor building. However, it does take a suction on the flash tank which is in turn connected to the Reactor Cooiant System vent header and to the letdown system. The vent header, letdown header and the Waste Gas System are not assumed to operate in any accident analysis. The letdown header and tne vent header are isolated by SFAS signals. The Waste Gas System is a nonessential system.

REACTOR BUILDING HEATING, VENTILATING AND AIR CONDITIONING SYSTEM Those portions of the Reactor Building Heating, Ventilating and Air Conditioning System associated with reactor building penetrations provide hydrogen purge and monitoring capabilities, containment to auxiliary building pressure equal:ation, and containment atmosphere punje.

The hydrogen purge ca3 abilities are included to allow venting of hydrogen produced by radiolytic reaction. See Section 2.1.5.a of this attachment for additional infonnation on this system.

Hydrogen created from this source would not become a problem for almost twenty days af ter an accident. Therefore this system is designed to isolate on a safety f eatures acuation' signal.

It can be returned to service for hydrogen monitoring and for hydrogen purge purposes when required.

It is classified as a nonessential system.

The reactor building containment air purge and equalization capability of this system are not used in any accident analysis. They must be isolated in accident situations. Therefore, they are classified as nonessential systens, and they are isolated by SFAS signals.

DEMINERALIZED WATER SYSTEM The Demineralized Water System penetrates the reactor building with a single line which branches to supply demineralized water to utility station hose conne tions, decontamination stations at the Fuel Transfer Canal, Reactor Head Storage Area and the West Decontamination Station, the Pressurizer Relief Tank and the Spray Lov at the Decontamination Station at the Reactor Head Storage Area. None <,f these functions is essential to accident mitigation; therefore the syste', is nonessential. The system penetration is isolated by a locked cle ed valve.

\\15\\

006 AUXILIARY STEAM SYSTEM The Auxiliary Steam System penetrates the Reactor Building Containment witn a line which brings steam to the pressurizer nitrogen pre heater. The preheater heats nitrogen gas which is used to provide a cover blanket for the pressurizer. The steam supply is not required to mitigate an accident; therefore the Auxiliary Steam System is non-essential. The system is f solated by a locked closed valve.

REACTOR COOLANT CHEMICAL ADDITION AND SAMPLING SYSTEM Chemical addition portions of this system are covered by the discussion in the section above titled Makeup and Purification Systen.

Sample system portions are not assumed to operate during accidents.

Lines f ron. the pressurizer, and from the pressurizer relief tank are automatically isolated by SFAS signals. Letdown sanple supplies are isolated by the letdown system SFAS valves.

Tne core flood tank sample and drain line is also nonessential.

It is isolated by a locked closed valve during nonnal operation.

PLANT AIR SYSTEM The Plant Air System supplies instrument air and other general air requi rement s.

The system penetrates *.he reactor building only to supply tne reactor building service air header.

The header is not used during plant operation and is isolated during plant operation by a locked closed manual valve.

C0OLANT RA0 WASTE SYSTEM The Coolant Radwaste System processes water letdown or drained from the Reactor Coolant System for boric acid recovery.

Connections with the letdown header and the reactor coolant vent header are described in the section or. the Waste Gas System. The reactor coolant drain header also interacts with the systea. The system is not assumed to operate in accident situations and it is thus nonessential. The reactor coolant drain neader is isolated by an SFAS signal.

COMPONENT COOLING WATER SYSTEM The Component Cooling Water System is not necessary f or saf e s. atdown and cooldown of the nuclear steam system.

It provides cooling water for reactor coolant pump seals and motors, f or the letdown coolers, and f or control rod drives. Operation of the system in accident situations may prevent damage to the reactor coolant pumps and control rod drives.

The system has no direct connection with systems containing radioactive fluids. The Component Cooling Water System is an essential system. The system has no direct connection with systems containing radioactive f luids. Tne system is not isolated by a SFAS signal.

1751 007 6

STEAM GENERATOR SYSTEM Once Through Steam Generators (OTSGs) offer an avenue for heat removal in accident situations. This heat removal function is an essential f unction which is not isolated by an SFAS signal. The system has no direct connection with any system containing radioactive fluids.

OTSG secondary sample and drain lines also penetrate the containment. They serve no accident mitigation purpose and are thus nonessential portions of the main steam system. They are isolated except during sampling and draining activities.

The OTSG acid cleaning line penetrates the containment.

It is not connected to the OTSG's during normal operation and it can communicate with the containment atmosphere. The acid cleaning line penetration is nonessential and is locked closed during normal operation.

AUXILIARY GAS SYSTEM The Auxiliary Gas System is not assumed to operate in accident situations. Thus the system is nonessential. The penetrations associated with core flood tank nitrogen, pressurizer relief tank nitrogen and pressurizer nitrogen are isolated by locked closed valves.

MISCELLANE0US LIQUID RADWASTE SYSTEM The Miscellaneous Liquid Radwaste need not operate in accident situations. Therefore, it is a nonessential system. One section of the system takes fluid from the B reactor building sump for nonnal containment water accumulation removal. The penetration associated with this function isolates on an SFAS signal.

2.1.5.a Dedicated Penetration f or cxternal Recombiner or Post-Accident Purge Systems Tne Rancho Seco Unit i licensing basis includes purge systems for post-accident comoustion gas control of the containment atmosphere.

The post-accident hydrogen purge system is described in the Rancho Seco Unit 1 FSAR (pages 5.2 - 33 and 34, Table 5.2 - 2, pages lA -

61, 9A - 2,14A - 28 and Appendix 14C).

The hydrogen purtje system uses redundant penetrations (52 and 53) for hydrogen control. Tha redundant penetrations, piping, valves and blowers meet single f ailure criteria for operatien of the system.

The isolation aspect of the system meets the requirements of Safety Guide 11 as discussed on page 1A - 79 of the FSAR.

1751 008 FSAR A,:pendix 14C demonstrates the capacity of the hydrogen puqe sy stem. Using Safety Guide 7 assumptions,18 scfm purge is required 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> after the accident to maintain hydrogen concentration at less than 4 pertent. The nominal capacity of each blower is 20 scfm.

As discussed on p.14A - 29 of the FSAR, the lines, valves and instrumentation associated with the hydrogen purge system 3re Class I.

Should a blower, which is not Class I, become inoperable following an accident, sufficient time exists to eneg ize the redundant blower, or to replace the defective blower (a spare is onsite as is aescribed in 14.5a - 15 of the FSAR).

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1751 009 To improve the isolation capabilities of the system the District is considering a change to the system configuration as shown in the figure above.

If this change is made, it would be completed during the 1981 refueling outage.

2.1.5.b Inerting BWR Containments Not applicable to Rancho Seco.

2.1.5.c Capability to Install Hydrogen Recombiner at Each Light Water Reactor Plant This requirement applies only to those plants eat include hydrogen recombiners as a design basis for licensing per the Dr. Denton's letter of October 30, 1979.

It is not applicable to Ranche Seco.

2.1. 6. a Integrity of Systems Outside Containment Likely to Contain Radioactive Materials (Engineered Saf ety Systems and Auxiliary Systems) for PWRs and BWRs.

Rancho Seco has undertaken a leakage reduction program which will encompass parts of the following systems:

High Pressure Injection, Seal Injection, and Make-up Systens Letdown System Decay Heat Removal System Reactor Building Spray System Coolant Radwaste System Miscellaneous Radwaste System Waste Gas System Reactor Coolant Chemical Addition and Sampling System Reactor Building Heating, Ventilating and Air Conditioning The portions of these systems outside of the containment which could contain high source level radioactive materials following an accident or serious transient are outlined below:

Hign Pressure Injection, Seal Injection, and Make-up Systems All segments of these systems outside of the containment are included 17 the leakage reduction program with the exception of those portions which are required f or boric acid addition and deionized water addition.

These sections are exempted from consideration because they could have no high source level water supply and because they are protected from inflow by system check valves.

1751 010

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es Letdown System Tne Letdown System outside of the containment is included in the leakage reduction program with the exception of system sections associated with boric acid and deionized water addition.

Oecay Heat Removal System The Decay Heat Removal System outside of the containment is included in the leakage reduction program with the exception of those portions of the suction header which draw f rom the borated water storage tank. No high source level flow would be expected into those portions of the system.

Reactor Building Spray System The portions of the Reactor Building Spray System outside the containment whicn draw suction from the decay heat suction header as well as the portions which discharge to the reactor building spray header are included in leakage reduction program. This specifically excludes the spray additive tanks and piping up to the spray additive euductor suction.

Coolant Radwaste System The Coolant Raawaste System would be automatically isolated from high source level activity.

However, to be responsive to the direction given by the D. G. Eisenhut letter of October 17, 1979, we include the flash tank, and tne reactor coolant system drain tank in the testing program.

These tanks will be tested using a helium leak or equivalent test.

Miscellaneous Liquid Raowaste System The portion of the Miscellaneous Liquid Radwaste System which lies upstream of both reactor building normal sump safety features isolation valves, SFV-66308 and SFV-66309 will be checked for leakage during nonnal operation. This is the only portion of the miscellaneous liquid radwaste system which might contain sigr,ificant amounts of high soume level fluid following an accident or serious transient.

Waste Gas System The Waste Gas System outside the containment, with the exception of liquid drains (normally closed), 7.nd sample lines (normally closed or remotely isolable), will be leak checked using helium leak detection or an equivalent detection metnod.

Reactor Coolant Cnemical Addition and Sampling System In section 2.1.8.a of this attacnment, the District presents an interim reactor coolant sampling system. That portion of the sampling system would be used during high source level situations.

Therefore, that portion of the system will be leak tested. The nonnal sampling system will not be tested because it will be isolated during high source level situations.

1751 011 Reactor Coolant Heating, Ventilating and Air Conditioning System Those portions of the system associated with the hydrogen purge system will be checked for leaks.

Portions of these systems have been visually inspected for evidence of leakage while in normal operation.

High Pra' ~ ire Injection, Seal Injection, Make-up System Letdown System Coolant Radwaste System Miscellaneous Liquid Radwaste System Results of this visual inspection indicate that valves listed below will oe worked to minimize leakage.

HV-22011 PLS-626 SFV-23616 PV-23606 FT-23606 Hi Root /Lo Root PSV-24014 SIM-012 SIM-106 SFV-23812 FT-23806 Hi Root /Lo Root FT-23808 Hi Root /Lo Root Reactor Coolant System allowable leakage is one gallon per minute for leaks af unknown origin and ten gallons per minute for leaks of known origin.

SP207.04A, Daily RCS Leakage Check, and SP207.048, Weekly RCS Leakage Test routinely monitor Reactor Coolant System leakage. By nature of the control volume boundaries, this test also checks large portions of the High Pressure Injection, Seal Injection, Make-up, and Letdown Systems. Thus, these routine surveillances will check or detect significant leakage changes in these systems.

The Decay Heat System Leakage Surveillance, SP203.09 will be perf ormed as usual during the 1980 refueling outage.

This surveillance procedure measures leakage on the decay heat system.

The Peactor Building Spray System will be inspected for evidence of leakage during performance of the quarterly reactor building spray system surveillance test, SP20a.03.

The Waste Gas System will leak tested during the refueling outage planned for January 1980. Tne test will employ a helium leak detection system.

1751 012 Two other systems which could contain radioactive fluids have not been included in this program. They are the Spent Fuel Cooling System and the Reactor Cealant System.

The Reactor Coolant System is not included because it is witMn tne reactor building. The Spent Fuel Cooling System will not be used in accident situations.

2.1.6.b Design Review of Plant Shielding and Environmental Qualificatien of Equipment for Spaces / Systems Whicn May be Used in Post Accident Situations.

The District has completed dose rate calculations during operation of the Decay Heat System, the Reactor Building Sp.ay System, the Coolant Sampling Systen, and the Letdown System. Fissica product soun:es were derived from cycle 4 depletion calculations. No isotopic decay was assumed. A system-by-system list of calculational assumptions fol lows :

Decay Heat System - The calculation assumed the system taking suction on the reactor coolant system and recirculating back to the reactor vessel.

It assumed no dilution beyond Reactor Coolant System mass based on water density c : Full power temperatures and pressures. Fission product :sotopic contribution included 100% of noble gases, 50% of halogens and 1% of Cesium and Rubidium.

Reactor Building Spray System - The calculation assumed the

. system was recirculating reactor building sump water.

It assumed the Reactor Coolant System water was diluted by the water in the borated water storage tank. Per the precedent set by Regulatory Guide 1.7, the calculation assumed no noble gas entrained within the depressurized sump water.

It assumea 50%

of the core halogen inventory and 1% of core Cesium and Rubidium i nventory.

Coolant Sampling System - The calculation assumed system flow was from the Letdown System to the primary sample sink and back to the Letdown System.

It assumed no dilution beyond the mass of the Reactor Coolant System based on water density at full power pressure and temperature. Fission product isotopic contributions included 100% of noble gases, 50% of halogens and 1% of Cesium and Rubidium.

Letdown System - The calculation assumed system flow from the Reactor Coolant System into the makeup tank.

It assumed no oilution beyond the mass of the Reactor Coolant System based on water density at full power pressure and temperature. Fission product contributions included 100% of noble gases, 50% of halogens and 1% of Cesium and Rubidium.

The results of the analyses to date show high dose rates in some areas of the auxiliary Du11 ding if all systems described above are operating. Much of the dose rate contribution would be eliminated by

\\15\\ DU Safety Features Acutation System containment isolation. Coolant Sampling System modifications described in section 2.1.8.a of this attachment will eliminate most Coolant Sartpling System dose rate contributions. Most importantly, the calculations show that the Control Room, the interim Technical Support Center and the Onsite Operational Support Center are acceptable for continuous occupancy during simutaneous operation of all of the systems described in the above aescription.

2.1. 7. a Auto Initiation of the Auxiliary Feedwater System See the District's November 19, 1979 letter to you dated.

2.1.7.b Auxiliary Feedwater Flow Indication to Steam Generator See the District's November 19, 1979 letter to you dated.

2.1. 8. a Improved Post-Accident Sampling Capability The District continues to work toward design of an interim liquid sampling system f or high sourte level measurements. We have chosen not to use dilution because of the complexity of the resulting system. We are considering the system shown in the figure below.

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The system will allow us to obtain a grab sample for analysis at a remote laboratory. The Discrict has identified a cumber of nearby government and private laboratories with the capability of handling and analyzing high level samples. The District has been and continues to confer with representatives from these laboratories to discuss terms and logistics associated with such a service.

The District will obtain containment atmosphere grab samples and analyze them with a similar arrangement.

For the long term the District intends to continue using remote laboratories for high level sample analysis. The District solicits NRC response to this approach.

1751 015 2.1.8.0 Increased Range of Radiation "oni '

The District divides commitments in this area into interim and long term comitments.

Interim commitments are t ose commitments which we intend to conplete prior to the end of tik. 1980 refueling outage.

Long term comitments are those which will not be implemented ~until the end of the 1981 refueling outage. Though the completion dates f or interim items have been extended to tne end of the refueling outage, we hope to have some of the interim items complete soon af ter mid January 1980.

High Range Noble Gas Effluent Monitors Comments in this area apriy to the reactor building purge vent, to the auxiliary bui'fing stack and to the radwaste service area vent.

Written procedures are being developed which will require in-situ readings using portable radiation monitoring instrumentation at roughly f fteen minute intervals. The individual perf onning the measurements will have verbal comunication capability with the control room.

In the long tenn, the District intendt to install one Eberline instrument corporation model SPING-4 system (or its equivalent). The instrument has a range of 10-7 to 105 micro Ci/cc for Xe - 133.

The instrument is a state-of-the-art instrument. We intend to install one instrument for eacn of the effluent paths described above.

The manuf acturer of the particular instrument described does not expect to complete qualification until November,1980.

The District will install instruments which satisfy the requirements of Regulatory Guide 1.97 Rev. 1, and ANSI N320-1979 specifications.

We do not believe it is prudent to commit to a regulatory document which has been distributed "for coment". Therefore we will consider satisfying tne requirements of Regulatory Guide 1.97 Rev. 2 when the approved document is distributed.

Post-Accident Radioiodine and Particulate Effluent Monitors The comments of this section apply to reactor building punje vents, the auxiliary building stack and the radwaste service area vent.

In the interim the existing lovl range vent monitoring instrumentation has grab sampling capability for both particulate and iodine collection. We are currently developing procedures which will address data analysis in the presence of high background levels and hign nobel gas levels.

In the long tenn a new high range nobel gas monitoring system will provide the capability to obtain particulate and radioiodine grab samples.

In addition the District is considering designation of a post-accident counting laboratory for analyzing such samples.

-'5-1751 016

In Containment Direct Radiation Level Monitor In the long tenn the District intends to purchase and install two Series 875 Victoreen grea monitors (or their equivalent) with a minimum range of 1-10'R/hr (photons). We intend to install the area monitors in seperate areas within the reactor building. The manuf acturer of the instrument described above is designing the instrument to comply with regulatory Guide 1.97 Rev. I and with ANSI N320-1979 specifications. The manuf acturer does not expect to qualify the instrumentation before June of 1980.

2.1.8.c In-plant Iodine Instrumentation The procedure has been uritten for radiciocine sampling using a low volume sampler with a silver zeoiite collection cartridge. The data analysis may be performed using either existing counting room equipment or portable monitoring instrumentation.

The procedure is currently undergoing the standard review process. A single channel analyzer is being purchased to aid in sample analysis. We expect to have the single channel analyzer in hand before startup af ter the 1980 ref ueling outage.

In the long tenn the District is considering designation of a post-accident counting laboratory for analyzing samples such as these.

2.1.9 Analysis of Design and Off-Nonnal transients See the District's November 26, 1979 letter to you. Also see Section 2.1.3.0 of this attachment.

2.2. l. a Shif t Supervisor Responsibilities The Rancho Seco administrative procedure AP-1, " Responsibilities and Authorities", has been rewritten to incorporate the shif t supervisor responsioility rcquirements of NUREG-0578.

fhe procedure is currently undergoing the procedural review process prior to final appro val. The review process should be complete by January 14, 1980.

2.2.1.b Snif t Technical Advisor See the District's November 26, 1979 letter to you.

The District wishes to clarify the conditions when the Shif t Technical Advisor will be assigned. The Shif t Technical Advisor will be assigned whenever the reactor is not in cold shutdown.

2.2.1.c Shif t and Relief Turnover Procedures Rancno Seco administrative procedure AP-23, " Control Room Watchstanding" has been rewritten to incorporate the shif t relief and turnover requirements of NUREG-0578. The procedure is currently undergoing the procedural review process prior to final approval.

The review process should be complete by January 14, 1980.

1751 017 2.2.2.a Control Room Access The Rancho Seco administrative procedure AP-23, " Control Room Watchstanding" has oeen rewritten to incorporate the control room access requirements of NUREG-0578. The procedure is currently undergoing the '1rocedural review process prior to final approval.

Tne review process should be complete by January 14, 1980.

2.2.2.b Onsite Technical Support Center The District has established an interim Technical Support Center in a former conference room adjacent to the control room. A window separates the conference room, here af ter called tne TSC, f rom the control room. The following drawing shows its relationship to the control room.

4!

y W

e i

CPERATIONAL SUPPORT CDTER i

i

\\

Q C

==

x COMPVILa LC00F.R pma c-3 l

' N

_ - ARCA/PFCCESS PF IATIC't I

/

i e re:I: me Pisrt

• 3,ylf7

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'$T C'

_4

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..cc::Tr:r. ic:w.

c

~',

[~ ^

y

.JC A.. _.J C

\\

r

'Y The control room and the TSC have common air conditioning and airborn radiation monitoring systems.

Therefore, control room airborn radiation monitor will also protect the TSC. Radiation protection procedures for the control room will be equally applicable to the TSC.

1751 018 The TSC currently contains technical information for advisory support in accident situations.

Included in that support are procedural support items such as plant operating procedures, process standards, technical specifications, emergency procedures, emergency plans, and maintenance procedures. Also included are piping and instrumentation diagrams and electrical one-line diagrams. Supplementing those hard copy drawings are more than 8000 aperture cards (film cards) with a visual viewing device.

Prior to startup af ter the 1981 refueling outage, the TSC will have the following communication capabilities:

5 Pacific Telephone lines 1 Station telephone extension 1 10 button mini switchboard telephone UHF radio telephone (portable from Control Room) on the District's radio net I speaker monitoring the UHF cross band radio to Sacramento County OE0 1 open speaker monitoring all control room audio 1 open mike (handset) to Control Room 1 dedicated 3 point line to the primary (Ione) and alternate (Herald) Emergency Operations Centers 1 dedicated 2 point line to the Corporate Emergency Center 1 walkie talkie for the security radio net 1 NRC " hotline" (red phone) to NRC Region V and NRC Headquarters Those capabilities marked with an asterisk are currently installeu.

Individuals in the TSC will be able to obtain information from the control room via the control room to TSC devices described above.

In

~ ddition, plant parameters can be ootained by going to the terminals a

in the computer room which lies directly across from the control room.

The terminals provide access to infonnation useful in assessing accident situations.

See tne District's November 26, 1979 letter concerning the find TSC conf iguration.

1751 019

_is_

2.2.2.c Onsite Operational Support Center Rancho Seco currently has an onsite operational support center (OSC). The OJC is where operators assemble to support normal or emergency plant operations. The diagram in the previous section shows its location.

Additior.al help such as maintenance and health physics personnel will gather at the ASSEMBLY POINT as described in the Rancho Seco Emegency Plan.

Containment Pressure Monitor (NUREG-0578 implementing letter of September 13, 1979 encl. 3, item 3 (1))

The District intends to install a safety grade containment monitor during the refueling outage in 1981.

Pressure indication will meet the requirements of Regulatory Guide 1.97, Rev. 1.

Containment riydrogen Concentration Monitor (NUREG-0578 implementing letter of Septembe r 13,1979 encl. 3, item 3 (2))

District will install a safety grade containment hydrogen monitor which meets the requirements of Regulatory Guide 1.97, Rev. I during the 1981 cafueling outage.

Containment Water Level (NUREG-0578 implementing letter of September 13, 1979 encl. 3, item 3 (3))

The District will install a narrow range reactor building sump level instrument to monitor the nomal containment sump "B" level during the 1981 ref ueling outage. T;1e narrow range level instruments will meet the requirements of Regulatory Guide 1.89.

The District will also install wide range containment water level instruments wnich meet the requirements of Regulatory Guide 1.97, Rev.1. The instruments, whicn will be of appropriate range, will be installed during the 1981 refueling outage.

Reactor Coolant System Vents (NUREG-0578 implementing letter of September 13, 1979 incl. 4).

The District has received the Babcock and Wilcox high point vent analysis. We are proceeding with review of the analysis as described in the District's November 26,1979 to you.

1751 020 ATTACHMENT 2 69110G921-00 DECEUBER 1979 I

e I,1 ADEQUATE CORE C00LIC]G DECAY I! EAT pie 00 VAL SYSTED 00DE OF OPEi!AT!DE!

t 1751 021

BWNP-20006 (6-76)

~

BABCOCK & WILCOX L ita NUCllAA POWER GENtRADQN OfVISCM 86-53 8-TABLE OF CONTENTS / EFFECTIVE PAGE LIST SECTION TITLE PAGE 000. NO.

1.0 INTRODUCTION

3 86-1105508-00 4

86-1105508-00 2.0 ANALYSIS 5LMfARY FOR LOSS OF RCS LNVENTORY 4

86-1105508-00 S

86-1105508-00 6

86-1105508-00 7

86-1105508-00 8

86-liO5508-00 9

86-1105508-00 FIGURE 1 LOCA LIMITS POWER SHAPE - 6 FT PEAK 10 86-1105508-00 FIGURE 2 SADDLE SHAPE POWER CURVE-UhKUE PEAKS 11 86-1105508-00 FIGURE 3

' SADDLE SHAPE PCWER CURVE-EQUAL PEAKS 12 86-1105508-00 FIGURE 4 LOCA LIMITS POWER SHAPE - 10FT PEAK 13 86-1105508-00 FIGURE 5 SM E L EREAK PCWER SHAPE 14 86-1105508-00 FIGURE 6 RCS PPES5URE VS CORE EXIT FLUID TEMPERATURE FOR 1400 F CLAD TEMPERATURE LIMIT 15 86-1105508-00 TEMPERAWRE FOR 1800, EXIT FLUID RCS PRESSURE VS CORE FIGURE 7 F CLAD TEMPERATURE LIMIT 16 86-1105508-00 FIGURE 8 CORE EXIT FLUID TEMPERATURE INDICATION TO LIMIT CLAD TEMPERATURE 17 86-1105508-00 FIGURE 9 SOURCE RANGE TRACE FOLLOWLNG REACTOR TRIP (TYPICAL) 18 86-1105508-00 FIGURE 10 CORFELATICN BETWEEN SOURCE RANGE LEVEL AND EVENTS FOLLOWING TMI-2 SCRAM 3/28/79 19 86-1105508-00 FIGURE 11 PRIMARY SYSTEM FLOW 20 86-1105508-00 TABLE 1 SIM4ARY OF FOA4 LNPtfr 21 86-1105508-00 3.0 ANALYSIS SLNMARY FOR REFLTLING CONDITIONS 22 86-1105508-00 1751 022 PAGE DATE:

12 5_79

BWNP-20007 (6-76)

BABCOCK & WILCOX Mumsen NUCLEAR POWER GEMetADQN DW153cN 86-1 ss a-00

_ TECHNICAL. DOCUMENT

1.0 INTRODUCTION

The TMI-2 Lessons Learned Task Force Status Report, NUREG-0578, contains two sections addressing inadequate core cooling. First, Section 2.1.9 requires that Licensees provide the analysis, emergency procedures, and training needed to assure that the reactor operator can recogni:e and respond to conditions of inadequate core cooling. Secondly, Section 2.1.3 requires that:

" Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement those devices cited in the preceding section giving an unambiguous, easy-to-interpret indication of inadequate core cooling. A description of the functional design requirements for the system shall also be included.

A description 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."

In response to NUREG-0578, an extensive program for inadequate core cooling has been developed. Ths objectives of this program are as follows:

1.

Develop operating guidelines that will allow the reactor operator to recogni:e and respond to conditions of inadequate core cocling under the following conditions:

a.

Power Operation with portions of the core in DNB.

b.

Loss of RCS inventory without the reactor coolant pumps operating.

i c.

Loss of RCS inventory with the reactor coolant ps operating.

d.

Loss of the Decay Heat Re= oval Systes and Loss of RCS Inventory During Refueling Operations.

e.

Loss of natural circulation due to loss of heat sink.

2.

Provide recc=mendations for any additional instrumentation required to indicate inadequate core cooling under the conditions listed above.

Included with the reccamendations will be:

a.

A coscription of the functional design requirements for the additional instrumentation.

b.

A desedption of the Operating Guidelines to be used with the proposed equipment.

c.

A description of the analyses used in developing these guidelines.

d.

Installation schedules for additional instrt=entation (if required).

1751 023 DATE:

PicE 1:_g_,,

3

BWNP-20007 (6-76) 8ABCOCK & WILCOX mu 86-110SS08-00 TECHNICAL. DOCUMENT To date, operating guidelines and supportive analyses are complete for the inadequate core cocling conditions described in Item 1 above.

For the inadequate core cooling conditions examined herein, guidelines for operator action and a description of the plant behavior, for use in operator training sessions, have been prepared. This info m ation is provided in the revisions to Part I and II of the Small Break Operating Guidelines (References 3, 4 and S) in Part I and II of the Operating Instruction for a Loss of Decay Heat Removal System and in Part I and II of the Operating Instruction for Nomal Power Operation.

Supportive analyses and infomation relating to the expected behavior of the out of core detectors and loop flow measurements during inadequate core cooling conditions is provided in Section 2.0.

2.0 ANALYSIS SIM4ARY FOR LOSS OF RCS INVEh"IORY Guidelines for inadequate core cooling and a description of the plant behavior to support additional operator training are presented in Parts I and II of the Small 3reak Operating Guidelines. These guidelines are in part based on the operators ability to assess the transient.

Section 2.1 describes the ba is and results of an analysis perfomed to correlate fuel rod conditions based on the pressure and te=perature conditions of the RCS. This information provides a means to detect and to initiate corrective actions for an inadequate core cooling event.

In addition, Section 2.2 and 2.3 provide a qualitative assess =ent of the behavior of the source range out of core neutron detectors and loop flow measurements during periods of inadequate core cooling.

2.1 Correlation of Cladding Temperature to Reactor Coolant Pressure-Te=cerature Condition During the small break LOCA, without the reactor coolant pumps operating, core cooling is accomplished by keeping the core covered by a steam-water mixture. However, should the core uncover, the uncovered portion of the fuel rod would be cooled caly by the steam produced by boiling in the covered portion of the red. Under this situaticn, elevated cladding temperatures, which could result in cladding rupture and/or a signifi-cant production of hydrogen due to metal-water reaction, would result.

The inadequate core cooling guidelines have been developed to allow the operator to detemine if core uncovery has occurred and to define appropriate actions to prevent significant cladding damage and/or hydrogen generation.

The core exit themoccupies, which measure the core outlet fluid temperature, are the most direct indicators available to the operator for detemining the core status during a small break LOCA. If these themoccuples indicate superheated fluid conditions, core uncovery is in progress. This behavior allows an assessment of core cooling.

1751 024 DATE:

12-S-79 FAGE 4

BWNP-20007 (6-76)

BABCOCK & WILCOX "u"0 EI muctua powen GENERADoM OMSiCN 86-1105s08-00

_ TECHNICAL. DOCUMENT To develop operator guidelines, a series of computer calculatiens were perfe:med to develop a correlation between the measured core outlet fluid truperatures and the peak cladding temperature. Using the above correlation, various levels of inadequate core cooling were defined and appropriate operator acticns were developed.

The approach taken for this analysis was to non-mechanistically reduce the Reactor Coolant System Inventory in order to develop the correlation between clad temperature and outlet fluid temperature. Core decay heat, based on 1.2 times the 1971 ANS standard for infinite operation, at 200 seconds after scram was utili:ed for this evaluation.

Core uncovery for small breaks should not occur any earlier than 200 seconds; thus this assumption will maximi:e power and the peak cladding temperature in the uncovered portion of the fuel red. Five power shapes, given in Figures 1 through 5, were analy:ed to cover a reascnable spectrum of core conditicus and to ensure that an outlet fluid te=pera-ture indication used for cperator action would correlate to a peak cladding temperature less than a selected value. Radial peaking factors were chosen such that the mnimm local power was equal to the LOCA limit value.

1 The FOAM code was utili:ed to predict the peak cladding temperature and core exit fluid te=perature. Table 1 provides a summary of the cases analy:ed. A brief outline of the procedure utill:ed in the FOAM code is as felicws:

1.

Using the input total core power, axial power shape, system pressure, and solid water level, the core mixture height is determined. This mixture height is based en a radial peaking factor of 1.0 and reflects the average core swell level.

2.

Assuming that all decay heat is re=oved in the covered portion of the fuel rod, the core steaming is calculated. As with the core mixture level, the steaming rate is based on a radial peaking factor of 1.0.

3.

Using the averap core steaming rate, the fluid temperature, in the uncovered portion of the fuel rod, for the hot pin is computed.

This calculation uses the input radial peaking factor. In dete: mining the fluid temperature, as a function of elevation in the core, it is assumed that all the core energy is re=oved by the steam.

4 Using the core steaming rate and the local fluid properties in the uncovered portion of the' fuel rod, a surface hea1; transfer coefficient, based en the Dittus-Boelter correlation ~, is calculated.

5.

Steady-State, hot pin cladding temperatures are then dete=2ined based on the local fluid properties obtained by Step 3 and the surface heat transfer coefficient obtained by Step 4.

175' 025 DATE:

PAGE 5

34

BWNP-20007 (6-76)

BABCOCK & Wil.COX

,aa:Aa powen oewetanow omsen ss-110ss0s-00 TECHNICAL. DOCUMENT Figures 6 and 7 su=mari:e the results of the calculations perfomed for the five power shapes analy:ed. These curves correlatu the calculated core exit fluid te=peratures for peak cladding temperatures of 1400F and 1800F, respectively. From these results, a bounding set of curves,, shewn on Figure 8, was obtained for use in the operating guidelines.

The small break operating guidelines include a provision for prcmpt tripping of the RC pumps upon receipt of a low pressure ESFAS signal.

If the RC pu=ps cannot be tripped, core cooling will be provided by the centinued forced circulation of fluid throughout the RCS. There are two ways that inadequate core cooling can occur for a small break with the RC pu=ps operating. First, with the RC pumps operating, the fluid in the RCS can evolve to a very high void fraction. Should the RC pumps trip at a time when the system void fraction is greater than approximately 70%, the amount of water left in the RCS wculd not be sufficient to keep the core covered and an inadequate core cooling situ-ation may exist. For this situation, the analysis described in the previous paragraphs apply directly.

Seccndly, if little or no ECCS flow is provided to the RCS, the fluid being circulated by the RC pu=ps will eventually become superheated steam due to the continued energy addition to the fluid provided by the core decay heat. Under these circumstances, an inadequate core cooling situation will start to exist. Due to the force circulation of the superheated steam through the core under these conditions, even with only one RC pu=p cperating, the heat removal process is better than the steam cooling mode described for the pumps off situation.

Thus, the indications and operator responses detemined for no RC pumps cperating are apprcpriate for controlling an inadequate core cooling situation with the RC pumps operating.

2.2 Excere Neutren Detector Behavior The execre source range neutron detectors are available to provide indications of ancmalous incere behavior, although they cannot uniquely quantify inadequate core cooling. A departure freu expected response is anticipated for conditions that lead to inadequate core cooling.

Therefore, the behavior of the source range detectors may, in some instances, be used to confim other indicatiens of inadequate core cooling.

The behavior of the neutron flux following a reactor trip is monitored and recorded by the source range count rate inst:umentation following reactor trip. An example of this trace is presented in Figure 5.

Nomally the detector count rate falls at rates characteristic of the various mechanisms of neutron production that exist following the t-ip.

During a trip, the neutren flux undergoes a prompt decrease associated with the negative reactivity of the control rods. Following the prcmpt decrease the neutron flux decays with an 30 second period, characteristic of the decay of the longest-lived delayed neutron group. The neutron 1751 026 PAGE 6

DATE:

12-5-79

SWNP-20007 (6-76)

BA8 COCK & WILCOX NUcLEAa Powet GENERADeN olvtSION s6-11ossoa.co

. TECHNICAL. DOCUMENT flux continues to deay at this rate until it approaches the level produced by neutron sources and subcritical =ultiplication. Two types of neutron sources are i=portant in the determination of neutron level following delayed neutron decay, namely:

(1) Fixed startup sources (2) Natural sources The most important of the natural sources is the photoneutron production (y, n) resulting from the interactions of high energy fission product gammas with deuterium (D,0). The photoneutron level decreases consistent 38 and I.al40)-

with the decay of fission products (primarily Kr The source range detectors will respond to a decrease in water density through several mechanisms.

(1) Reduced water density will enhance neutron transmission frem core to detectors.

(2) Reduced water density will decrease the neutron sources (i.e.,

photoneutrons from the y, n reaction in D,0).

(3) The reduced water density will decrease tHe core multiplication factor due to the negative modeFator coefficient.

Scoping calculations with a 1-dimensional transport code have shown that the dominate effect is the improved neutron trans=ission fron. core to detector. Thus, the source range detector count rate will increase or the rate of decrease will be altered, depending on the magnitude of the change in water density, even though the core is becoming more subcritical and the photoneutron source strength is decreasing.

The source range detectors cannot unattiguously detect inadequate core cooling because voiding in different regicas of the core will have different effects on the excore flux levels. If the reactor coolant pumps are cperating while the primary system is partially voided, the steam voids are expected to be evenly distributed. Under these conditions, the source range detectors are expected to read a higher than normal count rate. If the reactor coolant ms are not operating, the steam and water will separate. In ceder for the core to be inadequately cooled, the water level must drop below the top of the core. When this happens the source. range detector count rate should increase. However, as the level continues to drop, the continued decrease in the quantity of availabla water could reduce the photoneutron producticn and subcritical multiplication to the point where the source range detector output could begin to decrease. Because of this complex behavior, the source range detector should only be used to cenfi:m other indications of in-adequate core cooling.

1751 027 DATE:

PAGE 7

12-5-79

BWNP-20007 (6-76)

BABCOCK & WILCOX NUMBER NUCLEAA PoWtt GENERADoM Olvt$1CN 86-1105508-00

_ _ TECHNICAL DOCUMENT A correlation has been made between the source range detector response and several key events that followed the WI-2 accident on March 28, 1979.

This correlation is shown in Figure 10. The following is a discussion of the significant events referenced to the source range detector behavior shown in Figure 10. As was discussed above, there is censiderable uncertainty in interpreting the behavior of the source range detectors. Any interpretation should therefore be used with caution.

1.

Time 0400 - The neutren power in the reactor core decreased rapidly to the source range, as is typical of re= tor trip.

2.

Time 0408 - E=ergency feedwater was established to the stear generators approximately 8 minutes after reactor trip. The PORV had stuck open, and it continued to relieve reactor coolant. During the first 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following reactor trip, high-pressure injection (HPI) was initiated autcmatically severs 1 times as system pressure decreased.

Each time the autcmatic system activated, the plant operators took manual control of the HPI system. Information on HPI flow rates and times of injection was not available and had to be inferred from makeup tank levels, cperator inte: views, and the ala:s printer.

3.

Time 0420 to 0540 - Reactor coolant pumps were circulating a saturated, two-pnase flow. The void fraction was increasing due to loss of coolant through the FORV. From the SR plot, circulating two-phase flow can be inferred frcm the noisy, gradually increasing signal prior to point A.

The noise in the signal is due to the turbulent action in the two-phase fluid. A decrease in moderator density results in an increase in SR level due to reduced attentuation. At 0514, the RC pumps in loop B were turned off by the operator, but no fuel damage is believed to have occurred during this time interval since calculations have shown that the circulating fluid from loop A provided adequate heat removal.

4.

Time 0540 - The RC pu=ps in loop A were turned off by the operator.

As a result, the flow d6 creased rapidly, with a corresponding separation of steam and water. The calculated water level was at the bottom of the core inlet pipes, which are 3 feet above the top of the active core.

The calculation was based en coolant quality just prior to trip and was inferred to censist of 30 to 50% voids. This inference is consistent with gentile tube flow measurements and source range data.

5.

Time 0540 to 0615 - The water level in the core gradually decreased between points 3 and'D. The change in slope of the SR detector level at point C was interpreted to indicate the start of detector uncovering.

This is supported by the reflux boiler calculation and the coolant loss through the PORV. During this time, the RCS was acting as a reflux boiler; that is, steam was being created in the core region, condensing in the steam generators, and returning to the core by the cold legs.

The return of cold water to the reactor vessel was verified by the subcooled te=peratures obsmed in the cold legs during this period.

Reactor ecolant continued to be lost frca the system thrcugh the PORV.

1751 028 DATE:

"^ca a

12-3-79

BWP-20007 (6-76)

~

BABCOCK & WILCOX MuCLEAa Powtk otMERADoM OtVt38oM TECHNICAL. DOCUMENT 86-110550s-00 6.

Time 0615 to 0654 - The block valve upstream of the PORV was closed at 0615, preventing further loss of reactor coolant. The core was approximately 50% uncovered at this point and remained near this level until 0654 During this time interval, system pressure increased rapidly from 620 to 2150 psig. System pressure was then manually regulated using the block valve.

7.

Time 0654 - Based on alam messages, it was cc.lcluded that RC pump 2-8 was started and ra either intemittently or continuously for approximately 19 minutes. The core coolant level increased with at most 2 to 3 feet of the core remaining uncovered. This inferred level is supported by scme incore themoccuple readings which came on scale and read below 700F. In addition, the SR levels from E to F indicate a rgid increase in coolant density.

8.

Time 0654 to 0724 - The o;en PCRV block valve (0713), core boil off, and the tumang off of RC pump 2-B dropped the reactor coolant level so that approximately 4 to 5 feet of the core were uncovered during this time period.

9.

Time 0724 - The alam printer indicated that high-pressure injection of about 1000 gpm was started and continued for about 2 minute-before the operator took control. After this time, HPI flow is uncertain but apparently was at least reduced in flow. During this period, the core was partially refilled until only 2 to 3 feet of the core was uncovered. The temperature in the peripheral incere themo-couples decreased rapidly to the 600-700F range.

10.

After 0724 - The water level in the core gradually increased with minor perturbations. This was detemined from some peripheral themoccupies that ca=e back on scale, indicating the temperature was below the saturation level of 600F. Core covering was further substantiated by the retum of the SR detector readings to corrected nomal shutdown levels. At about 0730 the PCRV was cloced as detemined by the RC pressure increase.

2.3 Behavier of Loop Flow Indication Gentile flow tubes are used to measure mass flow in each loop. For solid water conditions, the reactor coolant pumps will act as constant volume pumps with the mass flow changing as the density of the water varies. If steam voids begin to fem in the loop, the reactor coolant pumps will still act as constant volume ms with some degraded per-fomance. The femation of steam voids in the loop reduces the fluid density and consequently the = ass flow in the loop. For this two-phase flow condition, the indicated flow will no longer accurately represent the mass flow. Mcwever, the indicated flow will follow the trend of a decreasing measured flow with an increasing void fraction. Figure 11 is the measured loop flow during the TMI-2 accident. This curve illustrates the expected behavior of the measured locp flow durin2 two phase flow conditions with a gradually increashtg void fraction.

DATE:

PAGE 3,.9 g

1751 029

1 i

o FIGURE 1.

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Figure 6 RCS PRESSURE VS CORE EXIT Flul0 TEMPERATURE FOR 1400*F Cl.A0 TEMPERATURE LIMIT

~ ~ ~ ~

1300 REF. FIGURE 1 1200 REF. FIGURE 2 1100 REF. FIGURE 3 f

1000 3

RE,F. FIGUR 900 P 4

2 1

2 m

O 800 3

u 700 >

REF. FIGURE 5 600 500 1

400 1751 035 200 600 1000 1400 1800 2200 Pressure, psia DATE:

1ll-5-79 PAGE 15 SERIAL: 86-1105508-00

Figure 7 RCS PRESSURE VS COPE EXIT Fl.Ul0 TEMPERATURE FOR 1800*F CLAD TEMPERATURE LIMIT 1600 REF. FIGURE 1 REF. FIGURE 2 O

1500 REF. FIGURE 3 1400

.m J

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1751 036 700 200 600 1000 1400 1800 2200 Pressure, psis DATE: 10-3-79 PME 16 SERIAL: 36-1105508-00

Figure 8 CORE EXIT FLula TEMPERATURE IN0lCATION TO LIMIT CLAD TEMPERATURE 1200 1100 a

TCLAD LESS THAN 1800*F j

1000 900 2

E m

5 800 1

a TCLA0 LESS THAN 8400*F 700 600 500 400 l 200 600 1000 1400 1800 2200 1751 037 Pressure, psia DATE: 12-5-79 PAGE 17 SERLy.:

86-1105508-00

FIGURE 9.

SGURCE RANC-E TRACE FCLLCWING siF'R TRIP (TYPICAL) 4 70 -

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10 12 14 Time From Reacter Trip, Hours 1751 038 DATE: 12-5-79 PAGE 18 SEIIA : 36-1105508-00

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.g IlddM DATE: 12-5-79 PAGE 20 SERIA: 36-1105508-00

BWP-20007 (6-76)

BA8COCX~& WILCOX NUCEAR POWtt GENERADoM DivlSION TECHNICAL DOCUMENT 86-110550s-00 Table 1.

Summary of FCAM Input Incut parameter Description Core power 1.2 X ANS at 200 see for 2772 Wt Core hydraulics 177 FA core with 15 X 15 fuel Axial power shapes 5 shapes (Ref. Figures I through 5)

Initial core water level 2 through 10 feet Core pressures 600,1000,1400,1800, 2200 psia Core inlet enthalpy h,,g Radial peaking factor Based on LOCA limit for ~av4-m local pcwer 1751 041 12-5-79 21

BWNP-20007 (6-76)

~ BABCOCK & WILCOX NucLEAS POWER otNERAfloN OlVISloN

.. _.__ TECHNICAL DOCUMENT 86-110ss08-u 3.0 ANALYSIS SUhMARY FOR REFUELING CONDITIONS Parts I and II of the Inadequate Core Cooling Operating Guideline provide direction to the operator in the event of indequate core cooling during refueling. The guidelines are based, in pat, on the reaction time of the operators. Section 3.1 describes die basis and results of the calcu-lation to determine the maximum time the operator has to act and the

.tinimum flowrate of water to recover the core in the event that the decay heat system fails. Section 3.2 describes the basis and results of the calculation to detexmine the maxi =um time the operator has to act and the minimum flowrate of water to recover the core in the event that there is a loss of reactor coolant system (RCS) inventory.

3.1 Loss of Decay Heat System After a failure of the decay heat removal system (IHR), the water in the reactor vessel will begin to heatup.

If the DHR system cannot be regained for a perioe. of ti=e the coolant will begin 'to boil and the liquid volu=e above the core will slowly decreas e.

The time frem the failure of the decay heat system until the liquid level in the reactor vessel reaches the top of the core was calculated. That liquid level is censidered to provide inadequate core cooling. The assumptions made in the calculation were:

1.

Decay heat level is 1.0 ti=es ANS Standard 2.

Time frem last shutdcwn is 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 3.

Decay heat level is constant for the duration of the transient.

These assu~ptions provide a heat input to the coolant of 0.42% of full

vactor power.

Initial plant conditions were defined as follows:

1.

Reactor vessel head is re=oved 2.

Reactor vessel water level is at the RV ledge 3.

Reactor coolant te=perature is 140 F 4

Reactor vessel refueling seal.is in place.

These plant cenditions define the heat removal capability of the coolant in the reactor vessel. The heat input will cause the coolant to boil in 26.7 minutes. It will take another 107.3 minutes for the coolant to boil down to the top of the core. The total ti=e frem the loss of the DHR system until the core is inadequately cooled is 214 hours0.00248 days <br />0.0594 hours <br />3.53836e-4 weeks <br />8.1427e-5 months <br />.

1751 042 DATE:

12-5-79 22

BWP-20007 (6-76) 8ABCOCK & WILCOX NUCLEAA Powet otNERADON OmsloN

1. i"""

. TECHNICAL. DOCUMENT The rate that coolant is boiled off in the reactor vessel will set the minima flowrate required to recover the system. The coolant will boiloff at 81.6 gallons per minute. The decay heat pumps can each supply a nominal flowrate of 3000 gpm as makeup frcm the BWST. As shown in Figure 1, this flowrate will reflood the reactor vessel in about 3 minutes. An absolute minimum flowrate of 250 g;m should be provided in order to reflood the core within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

3.2 Loss of RCS Inventory Prior to removing the reactor vessel head, a loss of RCS inventory is covered in the small break operating guidelines. After the reactor vessel head has been removed, the RCS inventory is assumed to be reduced non-mechanistically. The RCS inventory will be reduced as if there were a break of equivalent area to the cross sectional area of one incore no:tle. The time from the initiation of the loss of RCS inventory until the liquid level in the reactor vessel reaches the top of the core was calculated. That liquid level is considered to provide inadequate core cooling.

Initial plant conditions were defined as follows:

1.

Reactar vessel head is removed 2.

Reactor vessel water level is at the RV ledge.

3.

Reactor coolant temperature is 140 F 4

Reactor vessel refueling seal is in place Applying these initial plant conditions to the assumed break area 2enerates a volu=entric flowrate out of the break of 0.094 ft.*/sec.

Knowing the dia=eter of the reactor vessel, this flowrate is transformed to a reactor vessel level decrease rate of 0.456 inches per minute.

At that rate the top of the core will be uncovered in 3.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The minimum flowrate required to recover the core will have two components.

They are the leak rate and the boiloff rate. The leak rate will have to be overceme regardless of the status of teh DHR system. The DHR system will continue to maintain system temperature as long as there is still flow through the DHR system. However, when RCS inventory reaches the bottem of the hot leg no:tles, the DHR system will no longer function, and reactor coolant te=perature will begin to rise. It will take a finite amount of time for the coolant to boil, but that time has been assumed to be :oro. It has also been assumed that the coolant will boil off at the same rate as in the loss of DHR sy stem case. The total coolant loss will be 123.3 gpm (42.2 + S1.6 Apm) with no DER system.

The decay heat ms can each supply a ncminal f1;wrate of 3000 gym as makeup from the BWST. As shown in Figura 2, this flowrate will reflood the reactor vessel in about 3 minutes, regardless of the status of the remainder of the DHR system.

17b)

PAdE DATE.

12-5-79 23

BWP-20007 (6-76)

BABCOCK & WILCOX NUCLEAR POWER otNERATIQN Olvl33CN TECHNICAL DOCUMENT 88-** 88 8~

If the IHR system has also been lose, ut absciute minimum flowrate of 300 gpm should be provided in order to reflood the core within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

If operator actions are initiated so that the DHR system is not lost, an absolute mini =um flowrate of 200 gpm should be provided in order to reflood the core within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

1751 044 DATE:

PAGE 12-5-79 24

5 4

O 2

3 E:

2 1

g 3

6 9

12 15 18 21 23 24 27 30 33 36 Flowrate (x10-2) (GP'd)

REFLOOD TIMES FOR A LOSS OF OHR SYSTEM Figure 1 1751 045 an-12 5 g PAGE 25 SERIAL: 36-1105508-00

~.

5 4

3 WITH CORE BOILING 5

E 3

[

WITHOUT CORE B0lLING 1

0-3 6

9 12 15 18 21 24 27 30 33 36 Flowrate (x10-2) (gpy)

REFLOOD TIMES FOR A LOSS OF RCS INVENTORY Figurej27 51 04 6 DArs: 10 3 79 PACE 26 SERIAL: 36-1105308-00

SWNP-20007 (6-76)

BABCOCK & WILCOX wucusa powen ceN5tADoM OM3loM 86-110ss08-00

._ _ _..___ TECHNICAL DOCUMENT 4.0 ANALYSIS

SUMMARY

FOR POWER OPERATION - DNB CONDITION The core themal-hydraulic study was perfomed to investigate the DNBR response to two postulated conditicats which could result in inadequate core cooling. These conditions are:

1) An undetected increase in the core radial peaking (increase in hot assembly power relative to core aversge), and

2) an undetected reduction in core coolant flow.

While either or both of these conditions can be postulated to occur, to a limited extent, no credible mechanism has been postulated which could cause sither of these events to proceed undetected to an extent such that core cooling effectiveness would be reduced sufficiently to result in a d?parture from nucleate boiling (DNB). However, in order to assess the potential for inadequate core cooling resulting from these postulated events the following study has been perfomed indapendently of *any evaluations regarding the credibility of the occurrences.

4.1 Analysis Assumoticns and Initial Conditions These analyses represent a "best estimate" calculation of the minimum departure frem nucleate boiling ratio (DNBR), with the exctption that the initial core power distribution selected for the analysis was a design distribution without uncertainties. This design peaking dis-tribution bounds (in terms of w4-- asse=bly relative power) typical core power distributions in operating plants. The power distribution was held constant for the reduction in coolant flow study, without consideration of moderator feedback effects which would tend to flatten assembly peaks as coolant voiding occurs. A sy= metrical axial power distribution was assumed as a conservative (for DNBR calculaticns) representation of actual core axial distributions. Initial reactor operating conditions were selected to represent the nominal cperation of a typical Babcock 5 Wilcox reactor rated at 2772 *t core power.

The initial reactor coolant system (RCS) flowrate selected for analysis was 108% of design (where design flow is 88,000 GPM/pu=p), which is representative of the win 4mm measured flowrate in operating B6W reactors.

The RCS pressure and inlet temperature (2200 PISA and 557.30F, respectively) represent nominal operation at this flowrate.

Core and fuel assembly models used for this evaluation were representa-tive of the nominal gecmetry, without application of tolerances and engineering hot channel factors.

Computer codes used for this analysis included LYNX 1 (1) and LYNX 2 (2).

LYNX 1 was used to model the core, on an asse=bly by assembly basis, while LYNX 2 was used to =odel an individual asse=bly on a subchannel basis.

1751 047 DATE:

PAGE 27 12-5-79 1

BWNP-20007 (6-76)

BABCOCX & WILCOX meu NUCIAA Pows CENSADoM omsCN 86-1105508-00

_ TECHNICAL DOCUMENT 4.2 Method of Analysis and Results

1) Increase in Core Radial Pesking Various power increases in the hot pin were investigated by increasing the hot bundle relative power while maintaining normali:ation of the power distribution across the core. The power gradient around the hot bundle was also maintained by increasing the bundle radial power factors for the seven surrounding bundles in an eighth-core symmetric model.

A symmetrical axial power distribution (1.4 cosine) was selected as a conservative representation of actual core axial power shapes.

Figure 1 shews the effect of increasing the bundle radial power on the minimum DNBR response of the hot pin. A bundle radial peak greater than 2.3 is required to achieve the BAN-2 CIF design limit of 1.30 and a bundle radial peak of 2.45 is required to achieve a DNBR of 1.00.

Figures 2 through 5 provide bund.'s radial pcwer distributions and coolant exit temperature distribut,ons for core codels with hot bundle 4

radial powers of 2.3 and 1.8.

It can be noted the hot bundle and its surrounding bundles have coolant exit te=peratuze of Tsat for the case with a hot bundle radial jewer of 2.3.

2) Reduction in Core Coolant Flow Various reductions in the nominal flowraterwere studied in conjuncticn with the "best esti= ate" bundle radial power distribution in Figure 6.

This distribution is designated "best estimate" but still possesses higher bundle peaks than a typical B6W 2772 MWt core.

Figure 7 shows the effect of reducing the nominal flowrate en the mini =um DNBR response of the hot pin. A flow reduction greater than 42% is necessary to cbtain minimum DNBR values below the BAN-2 QiF design limit of 1.30 and a flow reduction greater than 55% to exceed a mini =um DNBR of 1.00.

Figures 8 and 9 provide coolant exit te=perat for two flowrates, 83% and 55% of ncminal system flow. pre distributions It can be seen that a majority of the bundles in the core will experience exit coolant temperatures of T fo of ncminal, system Now.r an inadequate core cooling condition of 55%

3 Figure 10 has been provided.ro shcw the effect of bundle radial power en coolant exit. temperature for various flowrates. Bundle exit coolant temperatures are at Tsat for bundles with radial powers as low as 1.10.

Since the thez:moccupies may have measurement errors of 10 F or more, the 0

significant trend is the. flat te=perature response when T is reached.

sat 1751 048 DATE:

12-5-79 PAGE 23

FIGURE 1 MINIMUM CNER(BAW-2 CFF CCRRELATICN) AS A FUNCUCN CF THE HOT SUNCLE RADIAL PEAK 3.0

" East Est2 mate" Design 2772 MWt 2200 osia 100% Power 108% of 88,000 GPM/Pumo Tin = 557.26*F 8:

A 0

5a g

2.0 G

7 5$

1.5 1

E E

1.0 0.5 i

1.5 1.7 1.9 2.1 2.3 2.5 Hot Sundle Radial Peak 1751 049 CATE: 12-5 ~9 PAGE :9 SERIAL: 36-1105308-00

FIGURE 2 BUNDI.E RADIAL PCWER DISTRIBUTION N

-_. 1.6526 _

.2.0450...

.0.2090

.2.2915....2.2434 -..

.0.8344

.0.3194...

.1.4069 _..

2.1930 2.3000*

2.2199 1.3252 0.8430 0.5349 0.2385

's 2.2644 2.1169 1.0274 0.7067 0.5291 0.1774 1.3337 0.8092 0.2648 0.2872

• HOT BUNDLE s

1/8th CORE SYMfETRY 0.6009 0.4198 0.1412

's N

0.2145

'N N

1) ATE: 12-5-79 PAGE 30 SERIAL: 86-1105503-00 1751 050

FIGURE 3 COOLANT TEMPERARIRE DISTRIBUTION AT EXIT 2.3 HOT BUNDLE RADIAL MINLMJM DNBR %1.0

\\.

W/#/ ///H[ 7649m.4

/

638.4... 649 4 649.4 638.4..,.

609.4 579.6 569.7 s

/

649.4 649.4 649.4 629.7 604.7 587.9 571.5

^

~

/l///

f////,

/

,649.4

/649.4 617.4 597.9 587.3 568.2

/

632.6 604.8 574.3 573.7 1/Sth CORE SYlNETRY s

[:

T,,,

591.9 581.0 565.9 i

569.7 N

\\

'1751 051 DATE: 12-5-79 PAGE 31 SERIAL: 36-1105508-00

FIGURE 4 BUNDLE RADIAL POWER DISTRIB!JrION s

s 1.2933

. 1.6005 _

1.7933 1.7557 _..

1.0197 1.5047 0.3943 1.4065 1.7163 1.8000' 1.7373 1.3252 1.0283 0.7202 0.4238

=

1.7722 1.6567 1.2127 0.8920 0.7144 0.3627 1.3337 0.9945 0.4501 0.4725 HOT BUNDLE j

s 1/Sth CORE SYMETRY O.7862 0.6051 0.3265 s

s 0.3998 s

's

\\

\\

DATE:

12-5-79 PAGE 32 SERIAL: 86-1105508-00 1751 052

FIGURE 5 COOLANT TEMPERATdRE DISTRIBUTION AT EXIT 1.800 HOT BUNDLE RADIAL

\\

622.1.

635.9 644.7 645.0...

630.9 614.5 588.2 579 0 s

I 641.3 645.7 643.1 625.7 612.4 597.5 581.6 s

644.5 640.6 621.1 606.6 597.5 578.5 626.7 611.9 584.0 583.8 1/8th CORE SUNETRY s

600.9 590.9

$76.1 N

579.9

\\

\\

N 1751 053

.DATE: 12-5 9 PAGE 33 SERIAL: 86-1105508-00

FIGURE 6 "BEST ESTIMATE" BUNDLE RADIAL POWER DISTRIBITTICN N

N 1.0830 1.3402

.. 1.5017 1.4702 1.4069....

1.1282 0.6132 0.5028.. _.

x s

1.4372 1.5073 1.4548 1.3252 1.1368 0.8287 0.5323 N

s N

1.4840 1.3873 1.3212 1.0005 0.8229 0.4712 N

N 1.3337 1.1030 0.5586 0.5810 N

N 1/Sth CORE SYMMETRY 0.8947 0.7136 0.4350 DENOTES THE HOT BUNDLE N

N N

0.5083 s

N

\\

1-751 054 DATE: 12-5-79 PAGE 34 SERIAL: 86-1105508-00

FIGURE 7 MINIMUM ONBR (8AW-2 CHF CCRRELATION) AS A FUNCTICN OF PERCENT NCMINAL FLCW 3.0 2.5 8:m o

3 2.0 E

"Best Estimate" Design 2772 MWt 7

2200 Psia g

1.5 Tin = 557.26*F-100% Pcwer c:

i 1

1.0 0.5 0.0 i

50 60 70 80 90 100 Percent Ncminal Flcw (,5)

DATE:

12-5. 9 PXE 33 SERIAL: 36-1105508-00 1751 055

FIGURE 8 COOLANT TEMPERA 1TJRE DISTRIBLTTICN AT EXIT 83% NCNINAL SYSTEM FLCW

\\

621.9

. 633.6.... 644.6 643.8

, 642.0...,

629.5 600.5

, 591.0

\\

1 641.0 645.5 642.5 636.5 627.5 611.2 593.1 x.

s 644.4 639.4 636.7 621.3 611.3 589.5 637.1 626.9 595.7 595.6 1/Sth CCRE SYINETRY N

614.9 603.7 586.7

~

s

\\

591.0 s\\

DATE: 12-5-79 PME 36 SERIAL:

86-1105508-00 1751 056

FIGURE 9 COOLANT TEMPERA 1TJRE DISTRIBITTION AT EXIT SS% NCMINAL SYSTEM FLOW MINIMUM DNBR %1.0 648.9 _.,_

9 2

621.8 f/

///h ?/ O V/ b V/

605.3

///

/

' /b W 49.//

/.4

/

649.4 649 649 649.

632.7 607.3 Y

/

f/

/

649. /

f649

.649.

648.6 632.4 602.8 X

/

/

s

/

1/8th CORE SYMETRY 2$s$9M 614.0 611.1

/

//

638.9 622.1 598.9

's 604.8 s

N DATE: 12-5-J9 PAGE 37 SERIAL: 86-1105308-00 1751 057

FIGURE 10 COOLANT EXIT TEM *ERATURE AS A FUNCTICN OF PERCENT NCMINAL FLOW 650 TSAT

'~~

Bundle Radial Pcwer 640 1.51 630 1.33 Y

2 620 3

l 1.08 ej 610 5

0.82 5

600 58 590 Minimum DNER 1.3 0.44 580 E n m.um CNBR 1.0 570 i

P,

e i

50 60 70 80 90 100 Percent.Ncminal Flew, P.

1751 058 CAT :

12-5-79 PAGE 38 SERIAL: 86-1105508-00

BWP-20007 (6-76)

BABCOCK & WILCOX MUCLEAR POWit otNERADQM OlYt33CM 86-110s50s-00

....-_.. ---. TECHNICAL DOCUMENT REFERENCES 1.

B. M. Dunn, C. D. Norgan, and L. R. Cartin, Multinode Analysis of Core Flooding Line Break for B&W's 2568 MWt Internals Vent Valve Plants, BAW-10064, Babcock 5 WU cox, Lynchburg, Virginia October 1975.

2.

Babu ck & Wilcox Revisions to EETAl-B, a Computer Code for Nuclear Reactor Core Themal Analysis (IN-1445), BAW-10094, Babcock 6 Wilcox, Lynchburg, Virginia, April 1975 3.

Operating Guidelines for Small Breaks for Oconee 1, 2, 3; Three Mile Island-1, 2; Crystal River-3; and Rancho Seco, Emergency Operating Specification 69-1106001-00.

4.

Operating Guidelines for Small Breaks for Arkansas Nuclear One-1, Emergency Operating Specification 69-1106002-00.

5.

Operating Guidelines for Small Breaks for Davis-Besse-1, Emergency Operating Specification 69-1106003-00.

6.

B. R. Hao, J. M. Alcorn, LYNX 1, Reactor Fuel Assembly Themal Hydraulic Analysis Code, BAW-10129, Babcock and Wilcox, Lynchburg, Virginia October 1976.

7.

LYNX 2, Subchannel Themal-Hydraulic Analysis Program, BAW-10130, Babcock and Wilecx, Lynchburg, Virginia, October 1976.

1751 059 DATE:

12-5-79 PAGE 39

BWNP-20004 (6-76)

BABCOCK & WILCOX NUCLEAR Powet GENERATCN 08VISCN TECHNICAL. DOCUMENT EMERGENCY OPERATION SPECIFICATION 69 1106921 00 Doc. ID - Serial No., Revision No.

for INADEQUATE CORE CCOLING DECAY HEAT REMOVAL SYSTEM MODE OF OPERATION 1751 060 PAGE 1

BtNP-20005 (6-76)

BABCOCK & WILCOX NUCLfAt POWtt GENERAfiCN OlvtSION NUMIII RECORD OF REVISION 69-1106921-00 REY. No.

CHANGE SECT / PARA.

DESCRIPTION / CHANGE AUTHORIZATION 00 Original Issue - D. A. Beckner Customer Service PREPARED BY

/.

t 642# [ w h DATE

/d1 // c/ 7 9 Y

APPROVED BY SS-Sy, DATE APPROVED BY /l /d(Na:5eJ b u b m-16.

DATE 11//9 U4 LTitigf (Name)<

(Title) -

APPROVED BY

//lud/I

[ b /)2pf. @ [d. h. DATE

/

7

' /f-(Tit 16)

APPROVED BY b#

de, T u b b 4 DATE Il[ /S [7')

M iM T

i2/b/ h APPROVED BY DATE (Name)

(Titly) 1751 061 DATE:

12-12-79 PAGE 2

ATTACHMENT 3 000. NO. 5611055G5 01 AN ALYSIS SUMiY ARY LN SUPPOR" 0F INADEQUATE C01E COOLING GUIDELINES s

1751 062 Babecck & Wilcox

BWNP-20006 (6-76)

~~ BAliCOCK & WILCOX Numsta NUCLEAR PowtR GENttADoM OlvtSCM

_____ TABLE OF CONTENTS / EFFECTIVE PAGE LIST 69-1106921-M SECTION.

TITLE PAGE 00C. NO.

CASE 1:

Loss of Core Cooling During Heat Removal via Decay Heat Removal System Part I RCS Pressure Boundary Intact, Filled or Drained 1.0 SYMPTOMS 4

69-1106921-00 2.0 IMEDIATE ACTIONS 4

69-1106921-00 3.0 PRECAUTIONS 4

69-1106921-00 4.0 FOLLCWUP ACTION 4

69-1106921-00 5

69-1106921-00 Part II RCS Pressure Soundary Not Intact, RV Heat Detensioned 1.0 SYMPTOMS 5

69-1106921-00 2.0 IMEDIATE ACTIONS S

69-1106921-00 3.0 PRECAUTIONS 5

69-1106921-00 4.0 FOLLOWUP ACTIONS S

69-1106921-00 6

69-1106921-00 Part III RV Heat Removed, Fuel Transfer Canal Filled or Empty 1.0 SYMPTOMS 6

69-1106921-00 2.O IMEDIATE ACTIONS 6

69-1106921-00 3.0 PRECAUTIONS 6

69-1106921-00 4.0 FOLLCWUP ACTIONS 6

69-1106921-00 7

69-1106921-00 CASE 2:

Loss of RCS Inventory During Heat Removal via Decay Heat Removal 1751 063 Part I RCS Pressure Boundary 5 tact or Head Detensioned 1.0 SYMPTOMS 3

69-1106921-00 2.O IMMEDIATE ACTIONS S

69-1106921-00 DATE:

12 12 7g PAGE 3

BWP-20006 (6-76)

BABCOCK & WILCOX wumsta NUCLEAR PQwtR GENERATCN ONISCN TABLE OF CONTENTS / EFFECTIVE PAGE LIST 69-1106921-00 SECTioll _ _

TITLE PAGE 00C. NO.

3.0 PRECAITTIONS S

69-1106921-00 4.0 FOLLOWUP ACTIONS 8

69-1106921-00 9

69-1106921-00 Part II Head Removed 1.0 SYMPTOMS 9

69-1106921-00 2.0 IMMEDIATE ACTIONS 9

69-1106921-00 3.0 PRECAttrIONS 9

69-1106921-00 4.0 FOLLOWUP ACTIONS 9

69-1106921-00 10 69-1106921-00 1.0 BASES 11 69-1106921-00 12 69-1106921-00 13 69-1106921-00 14 69-1106921-00 2.0 PLWT RESPONSE 15 69-1106921-00 16 69-1106921-00 17 69-1106921-00

3.0 REFERENCES

17 69-1106921-00 1751 064 DATE:

12,.12-79 PAGE 31

BWNP-20007 (6-76)

BX8 COCK & WILCOX NUCLEAR POWER otNSADQN Osvt34cN 69-1106921-00

.._._. TECHNICAL DOCUMENT CASE 1: Loss of Core Cooling During Heat Removal via Decay Heat Removal System I.

RCS Pressure Boundary Intact, Filled or Drained 1.0 SYMPTOMS 1.1 RCS temperature increasing.

1.2 DH system low flow alarm or DH pump tripped.

1.3 High DH cooler inlet or outlet temperature.

1.4 RCS pressure increasing.

1.5 Low suction pressure on DH removal pump.

2.0 IMMEDIATE ACTIONS 2.1 If DH pump is not running, determine cause and attempt to restart.

2.2 If DH pump is running, ensure adequate cooling water flow through DH cooler.

2.3 Stop any operations that would decrease RCS inventory.

3.0 PRECAUTIONS - Not Applicable 4.0 FOLLOWUP ACTION 4.1 If RCS pressure increases to maximum pressure for DH pump line per Plant Limits and Precautions, ensure drop line valves are shut.

4.2 Trip DH pu=p if suction pressure drops to minimum required p_er Plant Limits and Precautions.

4.3 Check for proper valve line up.

4.4 If the DH pu=p cannot be restarted, place alternate DH train pump and cooler in service and restore normal cooling.

4.S If normal DHR cannor be, restored and RCS is drained, refill RCS with MU/HPI pu=ps from BWST, then close RCS vents. Establish secondary cooling either with natural circulation or increase RCS pressure and start one RC pump. The atmospheric du=ps or turbine bypass valves, if there is a vacuum in main condenser, can be used.

1751 065 DATE:

12-12-79 PAGE 4

BWP-20007 (6-76)

BABCOCK & WILCOX NUCLfAR Powtt GENERATION DIVISION 8EI 69-1106921-00 TECHNICAL DOCUMENT 4.S.1 Ifsecondary cooling cannot be established immediately, an alter-nate flow path can be established to give more time until a DH train or secondary cooling can be placed into service.

4.5.1.1 Refill and pressuri:e RCS to 200 psig and letdown via the DHRS drop line, use a flow path through one idle DHR pump and its cooler to the MU system cross connect valve, through this valve to the suction of the MU pump, and then into the RCS via normal MU line.

CAttrION: This method will not be adequate to keep the core cool in all decay heat load conditions, so a DHRS should be returned to service or secondary cooling should be established as soon as possible.

NOTE: Do not overpressuri:er the DH drop line.

II.

RCS Pressure Boundary Not Intact, RV Head Detensioned 1.0 SYMPTOMS 1.1 Low DH pump suction pressure.

1.2 DH system low flow alars or DH pump tripped.

1.3 High DH cooler inlet or outlet temperature.

2.0 IWEDIATE ACTIONS 2.1 If DH pump is not running, determine cause and atte=pt to restart.

2.2 Ensure proper cooling flow through DH cooler.

3.0 PRECAIIIIONS - Not Applicable 4.0 FOLLOWUP ACTIONS 4.1 Trip DH pump _if suction pressure drops to minimum required per Plant Limits and Precautions.

4.2 Check for proper valve line up.

4.3 If the DH pump cannot be restarted, place alternate DH train pu=p and cooler in service and restore normal cooling.

1751 066 DATE:

12-12-79 PAGE 5

BWNP-20007 (6-76)

BABCOCK & WILCOX NUCLEAR power GENERAT1oM OlvisioN TECHNICAL. DOCUMENT 69-1106921-00 4.4 If the DHRS cannot be restored, throttle MU/HPI pump discharge and open MU/HPI pump st.ction from BWST. Start MU/HPI pump and open discharge slowly until water is observed coming from RV flange. Close DH drop line valves.

NOTE: This m:ct cause flow into incore tank.

Ensure watertight door is shut.

4.S Re-establish containment integrity.

4.6 Increase MU flow as necessary to keep core cool. Verify water vs.

steam flow from RV flange.

4.7 Restsre a DHR train to service as soon as pc Jible.

III.

RV Head Removed, Fuel Transfer Canal Filled or Empty 1.0 SYMPTOMS 1.1 DH system low flow alarm or DH pump tripped.

1.2 High DH cooler inlet or outlet te=perature.

1.3 Low DH pump suction pressure.

2.0 IjiMEDIATE ACTIONS 2.1 If DH pu=p is not running, determine cause and atte=pt to restart.

2.2 If cooler outlet temperature high, increase cooling water flow through DH cooler.

3.0 PRECAtJrIONS - Not Applicable 4.0 FOLLOWUP ACTIONS 4.1 Trip DH pump if suction pressure drops to minimum required per Plant Limits and Precautions.

4.2 Check for proper valve line up.

4.3 If the DH pump cannot be restarted, place alternate DH train, pump and cooler, in service.

1751 067 DATE:

.12-12-79 PAGE 6

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BABCOCK & WRf 0X Num:En NUCLEAR POWER oENetAfloN DivtSioN

-. TECHNICAL DOCUMENT 69-12 6" t-00 4.4 If the DH system cannot be restored and the seal plate is not installed, if possible, use bleed transfer pump to maintain core covered until seal plate is installed.

If bleed transfer pump cannot be used, throttle MU/HPI pump discharge and open suction from BWST. Use MU/HPI pump to maintain core covered.

CAUTION.

Boron concentration in bleed hold tank may limit use of bleed transfer pump for makeup.

4.5 Re-establish containment integrity.

4.6 When seal plate is installed, throttle MU/HPI pump discharge and open suction from BWST.

4.7 Start MU/HPI punp.

NOTE: Visusily insure that coze is covered with water.

Increase flow if steam i obserted in the core and allow overflow into refueling canal.

4.8 Return DHRS to service as soon as possible.

4.9 If DH system operation cannot be restored before BNST water is exhausted and canal is full, use spent fuel system coolers to ecol and recirculate refueling canal via fuel transfer tube.

NOTE: Flanges have to be removed from fuel transfer tube.

1751 068 DATE:

12-12-79 PAGE 7

BWP-20007 (6-76)

BABCOCK & WILCOX NUCLEAA POWER GENERADoM OlVl$loN 68-1 6921-00 LECHNICAL DOCUMENT Case 2: Loss of RCS Inventory During Heat Removal via Decay Heat Removal System I.

RCS Pressure Boundary Intact or Head Detensioned 1.0 SYMPTOMS 1.1 Decreasing pressurizer (PR2R) level if RCS is filled.

1.2 High or erratic DH system flow.

1.3 DH system low flow alarm.

1.4 Increasing containment sump level.

1.5 Possible increase in containment radiation.

1.6 Possible increase in containment humidity.

1.7 Decreasing RCS pressure.

1.8 High activity in service water.

2.0 IMMEDIATE ACTIONS 2.1 Maintain RCS inventory with water from BWST and, if possible, with bleed transfer pumps.

2.2 Establish containment integrity.

2.3 Determine if the leak is in DHRS and, if possible, locate and isolate.

3.0 PRECALTTIONS - Not Applicable 4.0 FOLLOWUP ACTIONS 4.1 Trip DH pump if suction pressure drops below mini =um pressure required by Plant Limits and Precautions.

4.2 If operating DHRS had to be removed from service to isolate leak, restore normal DH removal by placing alternate DH train in service.

1751 069 DATE:

12-12-79 PAGE 3

BWP-20007 (6-76)

~ ~ BA8 COCK & WILCoX NUCLEAR PoWEB GENERAT1 ore Olvt31oM 68-6921-00 TECHNICAL DOCUMENT 4.3 If the leak is in the RC System or cann'ot be isolated, start the other DH pump in the LPI mode to maintain, if filled, PR2R level or, if drained, water flow out the RV flange.

4.4 When RB sump has sufficient water for NPSH on LPI pump, suction can be switched to RB sump, only on pump that is in LPI mode.

II.

Head Removed 1.0 SYMPTOMS 1.1 Decreasing RCS level.

1.2 High or erratie DH system flow.

1.3 DH system low flow alarm.

1.4 Increasing containment sump level.

1.S Possible increase in containment humidity.

1.6 Possible increase in radiation level.

1.7 H'.gh activity in service water.

2.0 IMMEDIATE ACTIONS 2.1 Maintain RCS itventory with water from BWST and, if possible, with bleed transfe: pu=ps.

2.2 Establish containment integrity.

2.3 Determine i.f the leak is in DHRS and, if possible, locate and isolate.

3.0 PRECAUTIONS - Not Applicable 4.0 FOLLOWUP ACTIONS 4.1 Trip DH pu=p if suction pressure drops below minimum pressure required by Plant Limits and Precautions.

4.2 If operating DHRS had to be removed from service to isolate leak, restore normal DH removal by placing alternate DH train in service.

1751 070 DATE:

12-12-79 PAGE 9

BWNP-20007 (6-76)

~ BABCOCK & WILCOX NUCLEAR PCwtf GfMERATION Olvt33CN TECHNICAL DOCUMENT 69-1106921-00 4.3 If the leak is ih the RC System or cannot be isolated, maintain removal of DH with one DH train.

4.4 Maintain RCS level with water from BWST or, if possible, with bleed transfer pu=ps until the seal plate is installed, fuel transfer tube flange can be removed, and fuel transfer tube flange drains are shut.

4.5 Flood refueling car.a1 from BWST.

4.6 If leak is largo enough to maintain water in RB sump for a NPSH for DH/LPI pu:tp, start other DH/LPI pump in recirculation mode when sufficient vater is in RB su=p.

4.7 If leak is not large enough to maintain NPSH for a DH pump, the other DH pump will have to be cycled as necessary to maintain water level in refueling canal above mini =um required for fuel movement.

CAIJDION: Additional borated water sources may be required to maintain refueling canal level and provide a NPSH in RB sump for DHR pump. Do not allow DH pump to cavitate.

4.8 When it is determined that refueling canal level can be maintained, both fuel transfer tube valves can be opened for fuel movement.

\\15\\ DD DATE:

12-12-73 PAGE 10

BWNP-20007 (6-76)

BABCOCK & WILCOX NUCLEAR POWER Gar 45tATioN DivtSloN TECHNICAL DOCUMENT 69-11 6921-00

1. 0

• BASES The decay heat removal (DHR) system is placed in operation at primary system conditions of about 300 psig and 280F.

The DHR system will remain in operation throughout the refueling period until the reactor vessel (RV) head is reinstalled, the reactor coolant system (RCS) is repressuri:ed and a heatup is started.

Plant conditions during refueling can be divided into three basic categories. They are:

State 1: RCS pressure boundary intact, filled or drained; DHR system operating.

State 2: RCS pressure boundary not intact, RV head off or detensioned; RV water level at or below flange.

State 3: RV head off; fuel transfer canal filled.

Inadequate core cooling can result while in any of the three plant conditions; State 2 provides the operator with the least time for action. One of two occurrances will be the cause of the inadequate core cooling. First is a loss of the DHR system.

Second is a leakage of RCS 2.nventory due to break.

The loss of RCS inventory will cause inadequate core cooling only if RV water level is decreased belo's the botton of the het leg noz les.

This results in a loss of the DER systen by uncovering the DHR system suction line. A loss of the DHR system will allow reactor coolant (RC) temperature to increase.

If a leak is below the core, continued leakage will permit core uncovering.

If the plant is in State 1, the coolant will boil and the core may go thrcugh a DNB transient.

If the plant is in State 2, the coolant will boil off until the RV water level is at the top of the core, at which point inadequate core cooling exists.

If the plant is in State 3, the coolant te=perature, will rise locally, but bulk boiling in the fuel transfer canal is not probable.

The following sections describe the conditions which could cause inadequate core cooling during refueling, the action necessary to recover, the reasons for that action, and the instrumentation necessary to detect potential inadequate core cooling.

Loss of the DHR system and loss of RCS inventory are addressed separately.

During any of the transients, THE OPERATORS PRIME CONCERN MUST BE TO :GEP THE CORE COVERED WITH CCOLLNT.

1.1 Loss of Decav Heat Removal System There are several conditions which may caust-the DER system to be lost.

They are:

1751 072 DATE:

12-12-79 PAGE n

S BWNP-20007 (6-76)

~ BABCOOC &' WILCOX mu wucsAn rowse casatanon omscN 69-1106921-00 TECMICAL DOCUMENT A.

DH letdown suction valves fail shut.

B.

DH cooler discharge valves fail shut.

C.

DH pu=p fails.

D.

Loss of cooling water to DH cooler.

E.

Lcss of liquid level due to improper draining of the system.

F.

Inappropriate valve line up with other system (e.g., makeup and purification (MUSP), sampling, spent fuel cooling (SFC),

vents on coolers and pumps).

G.

Decay Heat system relief valve lifts but does not reseat.

H.

Inadvertant opening of dump to sump valves (not on all plants).

I.

A leakage of coolant either from the primary loop or attached

' piping which drains the RCS below the hot leg DH suction nozzle (this is a special case and is addressed in Section S.2).

1.1.1 Actions for State 1: A loss of the DHR system can be countered by either using the redundant train of the DER system or by shut-ting the DH letdown suction valve and utilising once-through steam generator (OTSG) cooling. Ensure any planned draining is isolated.

If the RCS is partially drained, the system will have to be refilled in order to utili:e OTSG cooling.

1.1.2 Actions for State 2: The DHR system may be lost through either of two metnods. For the loss of the DER system which is due to a loss of inventory (i.e., Items E through H), secure any plant operations in progress which might have caused the loss of inventory and isolate the faulty component. Once those operations are se-cured, the DH system will be functional again and the RV can be refilled from the borated water storage tank (BNST).

For a loss of the DHR system which is due to a loss of heat removal ability, (i.e., Items A through D), secure the operating DH pump and realign the DH system so that the unaffected DH train will provide core cooling.

(If not secured within five minutes of a loss of suction, the cumo will fail.) For tne worst case (decay heat production 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown and water level at the RV nostle belt) these actions must be completed within 15 minutes of the initiation of the event in order to prevent core boiling.

If the initial RV water level is at the RV flange, boiling will start about 26 minutes with a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown decay heat.

1751 073 DATE:

n.12 79 PAGE 22

4 BWP-20007 (6-76)

BABCOCK & WILCOX Numsen NUCLEAR power GENetAUoN OWISioN TECHNICAL DOCUMENT 69-1106921-00 If the DH system is lost due to a failure of the letdown suction valve, there is no way to retain a closed DH system.

Both the BWST and the spent fuel pool are available sources of coolant.

That coolant will have to be pu= ped into the core and allowed to fill the fuel transfer canal.

The BWST is the prefered source of water to be used, and is pumped into the fuel transfer canal through the RV by a DH or HPI pump.

When the fuel transfer canal is full, use one SFC pump to pu=p coolant from the fuel transfer canal to the DH recirculation and test line and into the RV.

The system will use one SFC pump and cooler to remove DH until the DH litdown suction valve can be repaired. The spent fuel pool is a secondary source of coolant, but it can only be recycled with the fuel transfer tube and fuel tilt pit gate open.

1.1.3 Actions for State 3: The core cannot be inadequately cooled.

The redundancy in the safety grade DHR system will provide core cooling for all cases except a failed shut letdown suction valve.

In this event, cooling will be through the cross connect with the SFC system.

1.1.4 There are several instrumentation devices which will provide in-dication of a loss of heat removal ability. The primary indi-cation of a loss of DH cooling are DHR system flow, DH pu=p suction pressure (if available), DH pump discharge pressure (if available) and DH pu=p differential pressure (if available). Other indica-tions of the loss are DH pump motor current (if available),

audible sounds of DH pu=p cavitation, DH cooler inlet or outlet te=perature, valve position indication, reactor building (RB) sump level, and incontainment radiation monitors.

1.2 Loss of RCS Inventory While the plant is being cooled by the DHR system, a loss of RCS inventory will cause inadequate core cooling only if the leak will allow draining the system to a level where the DH pu=p loses suction.

If the leak is below the core (for example:

incore no:tles) continued leakage can uncover the core.

1.2.1 Actions for State 1: Operator action to recover from these conditions is provided in Appendix A of the Small 3reak Operating-Guidelines. Those guidelines require coolant injection from the BWST using the LPI system or the HPI system (throttled) for coolant injection depending on the RCS pressure.

1751 074 DATE:

.12.12-79 PAGE 23

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BABCOCK & WILCOX wuct 4a mwen otNemanow omscu TECHNICAL DOCUMENT 69-1106921-00 1.2.2 Actions for States 2 or 3: A leak in the DH system or the incore no::les can reduce the fluid level below the DH suction no::les co coolant =ust be added to the system. The analysis and recom-mendations assume that inventory will be reduced as if there were a break of equivalent si:e to the cross sectional area of one incore no::le. While in State 2, the break area will cause a maximum outflow from the system of 42 gallons per minute. The outflow may be up to 84 gpm while in State 3.

The State 2 outflow is equivalent to a RV level decrease of less than one-half inch per minute. At this rate, it will take over 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> for the system to be drained down to the RV outlet no::le. Action should be taken to investigate the cause of the loss of level as soon as possible, and action must be taken within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the break initiation, if the initisi RV water level were at the RV flange. This time may be reduced to as low as 15 minutes when the initial water level is at the RV no::le belt as would be the case for maintanence in the steam generator lower head region. Water must be added to the system to ma.ntain the RV water level above the RV outlet no::.les so that the DH-system can remain functional.

If actions to refill the RV are delayed, indications to confirm the loss of inventory may become more noticable. The RB sump level alarm will activate indicating the presence of water there. DH cooler outlet temperature may increase due to the heat capacity loss of the discharged coolant.

But the primary temperature may be reduced by increasing cooling water to the DH cooler. The incontainment radiation monitors may also provide an indication that RV water level is decreasing.

1.2.3 There will be a noticable increase only when RV water level is such that the radiation levels from the core are approximately equal to the other radiation sources in the containment.

1.3 Soecial Precautions Several special precautions must be observed by the operator while the RCS is experiencing these transient conditions. They are:

A.

The DH pump must be secured within five minutes of a loss of pump suction to preclude failure.

3.

The DH pu=p must be secured and the DH system letdown suction valves shut if the RCS repressuri:es to a pressure greater than DHR system design pres,sure.

C.

Ensure that the DH suction no::le remains convered to prevent loss of the DH pump and subsequent loss of core cooling.

1751 075 DATE:

22-12-79 PAGE 14

BWP-20007 (6-76)

BAECOCr& WILCOX "umsen sucura rown censtanow omsion 69-1 6921-00 TECHNICAL DOCUMENT 2.0 PLANT RESPONSE The RCS response will be presented for both a loss of DER system and a loss of RCS inventory in each plant condition. The plant response, both with and without operator action will be discussed.

2.1 Loss of Decay Heat Removal System When heat removal ability is lost, the coolant which is in the RV will heat up.

With no operator action, the coolant will con-tinue to increase in te=perature until it reaches the boiling point. If the plant is in State 1, the plant will go through a DNB transient, and operator action will be required to restore core cooling.

If the plant is in State 2 with initial RV water level at the RV flange the water in the RV will begin boiling in about 26 minutes. The coolant will continue to boil off until it reaches the top of the core in about 2-1/4 hours. At this point the core is considered to be inadequately cooled and core damage may occur. At lower initial RV water levels the coolant will reach the toiling point faster and although the coolant will boil off at the same rate, it will take less time for coolant level to reach the top of the core due to decreased system volume.

If the plant is in State 3, the coolant temperature in the RV will begin to increase, but the large volume of water in the fuel transfer canal will prevent bulk boiling in the system.

Operator action is required in all states in order to prevent core damage. While in State 1, natural circulation will begin, but if the OTSGs are not in operation the coolant will continue to heatup until it reaches the saturation temperature for the system pressure. The operator must either restore the DH system or initiate OTSG cooling to prevent core damage.

If the condenser is not available, maintain secondary steam pressure with the atmospheric dump valves.

If the RCS is partially drained, the operator =ust either restore the DH system or refill the system (using either LPI or HPI pu=p) and initiate OTSG cooling. The RCS may repressuri:e above the decay heat system design pressure.

The operator =ust insure that the DHR system remains isolated while RCS systea pressure is above the design pressure of the DHR system.

While in State 2, the operator must restore DH removal to prevent core damage. With RV water level at the RV flange, RC average temperature will increase from 140F to 180F within 15 minutes of the loss of the DH system. This time may be reduced to about 10 minutes should maintanence which requires a significantly reduced RV water level be.in progress.

If DH removal capability is re-covered by thon, the system can be safely cooled back down to 140F.

If the syst.em temperature reaches the boiling point, fluid will be boiled off at a maximum rate of 81 gallons per minute based on 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> DH levels. By restoring the DH system with a makeup flowrate greater than 81 gpm, the system will become subcooled 1751 076 DATE:

.12-12-79 PAGE

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BA8 COCK & WILCOX NUm8tt NUCLEAR POWER GENERATION QiVI3loN 69-1106921-00 TECHNICAL. DOCUMENT again. The higher the makeup flowrate, the more quickly the system can be returned to normal conditions. A minimum flowrate of 250 gpm is recommended as that flowrate will fill the RV from the top of the core to the ledge in about 50 minutes. A 3000 gpm flowrate will fill the RV in 3 minutes.

While in State 3, bulk boiling in the fuel transfer canal is not probable. However, the operator should attempt to restore the DH system or cross connect the SFC system to the DH system in order to remove DH.

2.2 Loss of RCS Inventory A loss of RCS inventory may occur within the reactor loop or in attached piping systems inside or outside the containment. The leak si:e is expected to be small because pressures are low and no large component failures are expected. To illustrate the effects of a leak a break in the reactor vessel of equivalent si:e to the cross sectional area of one incore no::le is shown.

While in State 1, the RCS pressure is low so the leak rate is low.

The most. evident reactor response will be a gradual reduction of hot leg water level (e.g., a 50 gpm leak will lower the water level about 1/2 inch per minute when water is in the vertical part of the hot leg).

If planned draining is occurring, the leak will be imperceptable by system level indication, but other symptoms will be gradually changing. Considerable time is available to take corrective actions at a comfortable rate.

If the reactor loop is full, several hours are required to uncover the core. The operators goal is to keep the core covered. This may be done by isolating the leak (if known) or by adding makeup to keep the DH suction and the core covered. The proper choice of actions is to keep the core cooled by adding makeup, then atte=pt to find the leak.

If the RCS is partially drained, the time to core uncove:y will be less. A leak rate of 50 gpm will lower the RV water level from the RV flange to the bottom of the hot leg in about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

DH pu=p suction will be lost when the bottom of the hot leg is uncovered and core boiling will begin about IS minutes later.

Repressuri:stion of che RCS by steam will occur if no vents are

'open.

The operators goals are the same as above.

While in State 2, the plant response will be similar to State 1 except it may take less time to uncover the DH suction. With the RV head removed, evaporation in combination with the leak will reduce water level from the RV flange to the bottom of the hot leg in about 2-1/2 hours. Operator action is required to replace the coolant discharged out the break so that the core is covered 1791 n77 DATE:

.12-12-79 PAGE

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BABCOCK & WILCOX III NUCLEAA POWER otNERAT1oM ofVI3loN 69-1 6921-00 TECHNICAL DOCUMENT and core damage can be prevented.

If the reduction in RCS inventory cannot be stopped, the operator must start filling the refueling canal with water from the BWST. Coolant will continue to be circu-lated through the core by one DH pump, while the other DH pu=p is filling the canal. Now coolant will flow out of the break at a somewhat greater rate due to the increased static head. As soon as the refueling canal is full, fuel transfer to the spent fuel pool should begin. More than one BWST tank volume may be required to complete the operation in order to minimize radiation exposure to personnel on the fuel handling bridge. Leaked coolant will collect in the RB sump and basement and should be pumped back to the fuel transfer canal by a DHR pump operated in the LPI mode.

While in State 3, the plant response will be extremely slow because of the large amount of water stored in the canal. The only possible large leak is via the vessel flange seal plate, but this cannot cause inadequate core cooling as the DH suction is not uncovered.

The opentor action is to recycle coolant to ensure that the core remains covered.

3.0 REFERENCES

The model plant used in developing Sections S and 6 was Arkansas Nuclear Cne. Each other plant has different cross connects between the DHR system and SFC system, but the operators objective is always the same. He must keep the core covered and attempt to attain coolant flow through the core.

The solutions presented for inadequate core cooling assume operation of certain essential equipment. For a loss of inventory, it is assumed that both the DH pumps and coolers are in operation along with both spent fuel pumps and coolers. For a loss of the DH 1

system, only the component causing the loss of the DH system is assumed to fail.

Existing instrumentation is also required to function. The essen-tial instrumentation is DH system flow and DH system temperature.

Additional instrumentation, which will provide the operator with the data to fully identify the problem, is valve position indication of motor operated valves in the DHR system, DH pu=p differential pressure, du=p-to-su=p flow (if capability is present), incontain-ment radiation monitors and RB su=p level.

1751 078 DATE:

12-12 ~9 PAGE 17

DN8 AT POWER 1751 079

1.0 Introduction and Stinmary NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Tem Recommendations" requireo an analysis of the symptoms of inadequate core cooling and required actions to restore core cooling. At a subsequent meeting with the staff on August 9,1979, the staff expanded the scope identified in NUREG 0578 to include in these investigations the potential for inadequate core cooling while operating at power. B&W has comple:ed its evaluations into the potential for inadequate core cooling at power and concluded that no additional procedures are necessary since, for the possibilities evaluated and discussed in this report, inadequate core cooling could only be obtained by the operator ignoring numerous already existing alams, or by the existence of substantial damager to the reactor internals.

2.0 Analysis The investigation into the potential for inadequate core cooling required an evaluation of potential means for the reactor to reach an inadequate core cooling state while maintaining 100% power. As a result of these evaluations, two potential modes of achieving inadequate core cooling at power have been identified as,1) an increase in core radial peaking, and

2) a decrease in core coolant flow.

2.1 Increased Radial Peakg In the case of increased radial peaking (increased power production from a local region of the core), the core exit themoccuples in the region of increased power generation would indicate an increasing trend, up to some temperature level representative of saturated coolant conditions. These conditions would also be indicated by core power distribution measurements, control rod position alarms, etc.

2.1.1 Analysis Two parallel analyses were perf omed to detemine how significant radial peaking increases could occur and to define the core themal-hydraulic response to various assumed radial peaking increases.

2.1.1.1 Themal-Hydraulic Analyses Analyses were performed to detemine the Departure from Nucleate Boiling Ratio (DNBR) response to increases in core radial peaking.

Starting from a nominal,100% power condition with a design peaking distribution, power in the hot assembly and the surrounding eight assemblies was assumed to increase, relative to the core 1751 080

average, until a ONBR of 1.0 was attained. The resulting hot assembly radial peaking f actor was 2.45 (that is, the power generated in the hot assemoly was 2.45 time that of an average assemoly).

2.1.1.2 Nuclear Analysis The FLAME code was used to investigate the radial peaking behavior of B&W cores for a number of typical f uel cycles. B&W plants nonnally have three banks of control rods (5, 6 and 7) with 8-12 rods each, designated as regulatory banks.

For each of the core types analyzed, the regulatory control rod banks were positioned at various locations throughout their full range of movement (0-300% withdrawn) and the resulting power distributions at 100% FP were recorded.

The maximum increase in tne radial peak above the nominal value for steady state operation was f ound to be 24% for any position in the 0-300%

range. However, the maximum increase before a rod positioq alann would be activated was found to be less than 3%. The maximum value of the assemoly radial peak for nominal steady state operation for B&W cores has been 1.55 for all past and present cycles.

If in addition it is postulated that a radial tilt is induced, the established relationship between tilt and peaking is a 1.5% increase in total power for a 1%

increase in tilt. The maximum peak increase prior to a tilt alarm in any B&W plant would be 7.5%, which could be assumed to be totally applied to the radial peak.

Thus within tne imbalance, rod position and tilt alarm limits the maximum anticipated assembly radial peak would be 1.55 x 1.03 x 1.075 = 1.72 and the maximum value with the control rods exceeding their alann limit would conservatively be 1.55 x 1.24 x 1.075 = 2.07.

Escalation of the indicated tilt toward its maximum allowable value within Technical Specifications before a significant power reduction is required could increase this conservation estimate of the radial peak value to 2.25.

However, at this point, at least two and possibly three alanns would be activated, and following the existing action statements of tne applicable Technical Specifications will ensure that no condition will be allowed where the assembly radial peak will 1751 081

exceed the DNBR limit. No steady-state condition was found within the RPS, for any amount of regulatory rod insertion, which would produce an assembly radial peak large enough (2.45) to result in an MDNBR 1.0.

For these reasons, no additional procedures or action statements are required to preclude DNB during core power ope ration.

2.1. 2 Plant Response to Increased Radial Peaking The response of the core to regulating bank motion is a change in the indicated power imbalance. As the rods are driven into the core from the nominal full power position ( 290% withdrawn for unrodded operation, 200% withdrawn for rodded operation), the imbalance will initially become increasingly negative and in mos* cases the RPS imbalance trip limit would be reached.

In order to stay at full power while this insertion was occurring, a continuous deboration would be necessary.

In most instances, operation at full power with bank 7 (unrodded) or banks 6 and 7 (rvdded) inserted could only be achieved without a trip by positioning the banks at low power and then deborating to increase the power level while holding the rods fixed.

In the latter case, the rod insertion alarm would be activated, while possibly the imbalance alarm would not (i.e., a low imbalance with deep rod insertion is possible).

In the case of rod insertion at full power, both rod insertion and imoalance alams would be activated after rod motion of 10-20% withdrawal.

The quadrant tilt alarm would be activated by any tilt-producing mechanism which produced a tilt larger than the allowable steady-state limit. The range of indicated tilt levels wnich would produce this alarm is approximately 2.4-4.0%, depending on the core and the age of the symmetric detectors.

In addition to the alarms discussed aoove, core exit thermocouples, called In-Core Themocouples (ICTC's),

would indicate a departure from nominal core operatio n.

ICTC's typically indicate a range of temperatures wh4 5 are related to core power or a localized increase in power di st ribution.

F generation, the corresponding ICTC's would indicate an increase in outlet temperatures, and, if tnis condition were severe enough to approach a DN8 situation, ICTC's in the region of increased power 1751 082

s s

would reach satiration temperature.

Three or more ICTC's, in close proximity to each other, having temperature levels significantly above normal would be indicative of a potential local peaking increase.

However, this situation would not be expected to occur without other " abnormal" indications, such as an indicated core power tilt or high indicated relative assembly power levels for the same core region.

2.2 Reduced Core Flow For the case of decreased core coolant flow, all core exit thermocouples would be expected to show an increasing trend, leveling out a value representative of saturated coolant conditions.

2.2.1 Analysi s A reduction in core cooling flow, with no corresponding reduction in Reactor Coolant System Flow, was postulated. Since RCS hot leg and cold leg temperatures would not be aff ected, this situation would not be detected by the Reactor Protection S yst em.

Analyses were perf ormed to determine the Departure from Nucl! ate Boiling Ratio (DNBR) response to reductions in core coolant flow. Starting from nominal, f ull-power, operating conditior,s tne core coolant flow was reduced until a DNBR = 1.0 was attained. The resulting core coolant flow was 45% of the initial value (a reduction of 59).

2.2.2 Plant Response The primary response to a reduction in core coolant flow would be an increase in the core exit temperature level. Since this could be postulated to occur as a result of the opening of an undetected core bypass flow path, RCS temperature (hot leg and cold leg) and flow measurements would not be expected to re spond.

However, in-core tnermocouples (ICTC's) would respond, with all working ICTC's showing increased (relative to nominal) temperatu res. A potentially severe situation, approacning DNB, would be indicated by ICTC's in high-powered core regions approaching saturation (sat).

3.0 Conclusions As a result of the evaluations described above, no substantive changes to existing operating procedures are necessary.

The investigations indicated that to obtain inadequate core cooling at power, the operators would need to ignore numberous already existing alarms, or major non-mechanistic damage to reactor internals would need to occur.

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b Operator Orientation for Inadequate Core Cooling due to DNB at Power 1.

Increased Radial Peaking causing DNB at Power Analyses show that Departure from Nucleate Boiling (DNB) does not occur until a radial peak of 2.45 occurs. A radial peak of 2.45 means that the hottest f uel assembly is generating 2.45 times more power than the average assembly. Analyses were then attempted to force a peak of this size by manipulating control rod position and power tilt. The maximum peak which occurred, while control roa position, power tilt, and imbalance were at Plant Technical Specifications limits was 1.72 and Reactor Protection System limits was 2.07 In both cases, this was f ar from the value required f or DNB.

The radial peaking necessary to reach DNB at power is clearly outside the limits for operation per Plant Technical Specifications and protective f unctions provided by the Reactor Protection System. Therefore, no additional operator instructions or guidelines.re necessary.

2.

Reduced Core Flow causing DNS at Power Analyses were performed to determine the amount of reduction necessary to cause Departure from Nucleate Boiling (DNB).

It was necessary to reduce total core flow to 457. of normal flow to get DNB at nonnal operating preuure and temperature. By the postulated increased core bypass flow method, the only direct indication of inadequate core cooling would be increasing core outlet thennocouple readings. But with the massive bypass flow nece;sary for DNB occurring, no further instructions would be valid for monitoring core exit thermocouples under this condit. ion.

The amount of reduced flow through the core necessary to have DNB is so large that it would take a major rearrangement of mechanical components inside the reactor vessel for this situation to occur.

Continued operation with this major rearrangement would be incredible. Since a smaller reduced flow through the core will not cause DNB, additional operator instructions or guidelines are not necessary.

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