Text

NRC Bulletin 1979-005A: Nuclear Incident at Three Mile Island - Supplement
	ML031210600
Person / Time
Site:	Three Mile Island
Issue date:	04/05/1979
From:	; NRC/IE
To:	;
References
	-RFPFR BL-79-005A, NUDOCS 7904170040
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__ -. S ORIGINAL-UNITED STATES NUCLEAR REGULATORY COMXISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 APRIL 5, 1979 IE Bulletin 79-OSA SUPPLEMENT NUCLEAR INCIDENT AT THREE MILE ISLAND Description of Circumstances:

XRC since issuance of IE Preliminary irformation received by the six potential human, 1, 1979 has identified in Bulletin 79-05 on Aprilfailures the core damage and which resulted design and mechanical The Mile Island Unit 2 nuclear plant.

radiation releases at the Three supplement clarify and extend the original information and actions in this chronology of the ThI accident Bulletin and transmit a preliminary (Enclosure 1.

through the first 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months of the event, loss of feedwater, both

1. At the time of the initiating out of service.

auxiliary feedwater trains were valved pressurizer electromatic relief valve, which opened during

2. The the pressure surge, failed to close when the initial pressure actuation level.

decreased below the pressurizer, the pressurizer 3; Following rapid depressurilzation of the inferences of high level indication may have lead to erroneous The pressurizer level indi:ation level in the reactor coolant system. pressure to prematurely terminate high led the operators substantial in the reactor apparently flow, even though voids existed injection coolant system.

isolate on high pressure injection

4. Because the containment does not water from the relief (HPI) initiation, the highly radioactive of the containment by the automatic valve discharge was pumped out This water entered the radioactive initiation of a transfer pump. auxiliary building where some of it vaste treatmentthesystem in the discharge overflowed to floor. Outgasslng from this water and wag system and filters thrcugh tie auxiliary building ventilation of radioactive noblt

the principal source of the offsite release gases.

system was Intermittently

5. Subsequently, the high pressure injection coolant inventory losses operated attempting to control primary apparently-based on through the electromatic relief valve, and 'r Due to the presence of steam pressurizer level indication. in the reactor coolant system.

noncondensible voids elsewhere in primary coolant inventory.

this led to a further reduction 79041 70 04.0

IE Bulletin 79-05A April 5, 1979 Page 2 of 5

6. Tripping of reactor coolant pumps during the course of the transient, to protect against pump damage due to pump vibration, led to fuel damage since voids in the reactor coolant system prevented natural circulation.

Actions To Be Taker. by Licensees:

For all Babcock and Wilcox pressurized water reactor facilities with an operating license (the actions specified below replace those specified in IE Bulletin 79-05):

1. (This item clarifies and expands upon item 1. of ZE Bulletin 79-05.)

In addition to the review of circumstances described in Enclosure I of IE Bulletin 79-05, review the enclosed preliminary chronology of the TMI-2 3/28/79 accident. This review should be directed toward understanding the sequence of events to ensure against such an accident at your facility(ies).

2. (This item clarifies and expands upon item 2. of IF Bulletin 79-05.)

Review any transients similar to the Davis Besse event (Enclosure 2 of It Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facillty(ies). If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safetv significance together with a description of any corrective actions taken. Reference may be made to previous information provided to the KRC, if appropriate, in responding to this Item.

3. (This item clarifies item 3. of IE Bulletin 79-OS.)

Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:

a. Recognition of the possibility of forming voids in the primary coolant system large enough'to compromise the core cooling capability, especially natural circulation capability.

b. Operator action required to prevent the formation of such voids.

c. Operator action required to enhance core cooling in the event such voids are formed.

IE Bulletin 79-OWA April 5, 1979 Page 3 of 5

4. (This item clarifies and expands upon item 4. of IE Bulletin 79-05.)

Review the actions directed by the operating procedures and training instructions to ensure that:

a. Operators do not override automatic actions of engineered safety features.

b. Operating. procedures currently, or are revised to, specify that if the high pressure injection (HP!) system has been automatically actuated because of low pressure condition, it must remain in operation until either:

(1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The MLI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure. If 50 degree subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.

c. Operating procedures currently, or are revised to, specify that in the event of HP1 initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain operating.

d. Operators are provided additional information and Instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system.

5. (This Item revises item 5. of IE Bulletin 79-05.)

Verify that emergency feedwater valves are in the open position in accordance with Item 8 below. Also, review all safety-related valve positions and positioning requirements to assure that valves are positioned (open or closed) in a manner to ensure the proper operation of ergireered safety features. Also review related procedures, such as those for maintenance and testing.

to ensure that such valves are returned to their correct positions following necessary manipulations.

ZE Bulletin 79-05A April 5, 1979 Page 4 of 5

6. Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to cause containment isolation-of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety injection.

7. For manual valves or manually-operated motor-driven valves which could defeat or compromise the flow of auxiliary feedwater to the steam generators, prepare and implement procedures which:

a. require that such valves be locked in their correct position; or

b. require other similar positive position controls.

8. Prepare and implement imediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 1002 flow capacity, are operable at any time when heat removal from the primary system is through the steam generators. When two Inde-pendent 100% capacity flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

When at least one 100l capacity flew poth is not available, the reactor shall be made subcritical within one hour and the facility placed in a shutdown cooling mode which does not rely on steam generators-for cooling within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or at the maximum safe shutdown rate.

9. (This item revises item 6 of IE Bulletin 79-05.)

Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently.

In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems red indicate:

a. Whether interlocks exist to prevent transfer when high radiation indication exists, and

b. Whether such &asters are isolated by the containment isolation signal.

-E Bu~letin 79-05A April 5, 1979 Page 5 of 5

10. Review and modify as necessary your maintenance and test procedures to ensure that they require:

a. Verification, by inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service.

b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.

c. A means cf notifying involved reactor operating personnel whenever a safety-related system is removed from and returned to service.

11. All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island U'nit 2 plant and ether actions taken during the early phases of the accident.

12. Review your prompt reporting procedures for KRC notification to assure very early notification of serious events.

For Babccck and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979. Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for response has not been changed.

Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979.

Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the KRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.

For all other reactors with an operating license or construction permit, this Bulletin Is for information purposes and no written response is required.

Approved by GAO, E 180225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems.

Enclosures:

1. Preliminary Chronology of THl-2 3/38179 Accident Until Core Cooling Restored.

2. Litt of IE Bulletins Issued in last 12 months.

IE Bulletin 79-05A April 5, 1979 PFFLIMINARY CHRONOLOGY OF TMI-2 3/28/79 ACCTDENT UNTIL CORE COOLING RESTORED TIME (Approximate) EVENT about 4 AM Loss of Condensate Pump (t a 0) Loss of Feedwater Turbine Trip t 6 sec. Electromatic relief valve opens (2255 psi) to relieve pressure in RCS t a 9-12 sec. Reactor trip on high RCS pressure (2355 psi) t a 12-15 sec. RCS pressure decays to 2205 psi (relief valve should have closed) t - 15 sec. RCS hot leg temperature peaks at 611 degrees F, 2147 psi (450 psi over saturaticn) t - 30 sec. All three auxiliary feedvater pu=ps running at pressure (Pumps 2A and 2B started at turbine trip). No flow was injected since discharge valves were closed.

t - mmin. Prersurizer level indication begins to rise rapidly t - a min. Steam Generators A and B secondary level very low - drying out over next couple of minutes.

ECCS initiation (HPI) at 1600 psi t a 2 min.

t a 4 - 11 min. Pressurizer level off scale - high - one HPI pump manually tripped at about 4 min.

30 sec. Second pump tripped at about 10 cin. 30 sec.

t a 6 min. RCS flashes as pressure bottoms cUt at 13' pasl (Mot leg temperature of 584 degrees F) t a 7 min., 30 sec. Reactor building sump puMp came on.

. . 0 TIMSE EVENT t - 8 mi. Auxiliary feedwater flow is initiated by opening closed valves t a 8 min. 18 sec. Steam Generator B pressure reached minhlur t a 8 min. 21 sec. Steam Generator A pressure starts to recover t - 11 min. Pressurizer level Indication comes back on scale and decreases t a 11-12 min. Makeup Pump (ECCS HPI flow) restarted by operators t a 15 min. - RC Drain/Quench Tank rupture disk blows at 190 psig (setpoint 200 psig) due to continued discharge of electromatic relief valve t - 20 - 60 min. System parameters stabilized in saturated condition at about 1015 pslg and about 550 degrees F.

t - 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 15 cin. Operator trips RC pumps in Loop B t - 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , 40 min. Operator trips RC pumps in Loop A t 3/4 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months CORE BEGINS HEAT UP TRANYSIENT - Hot leg temperature begins to rise to 62n degrees F (off scale within 14 minutes) and cold leg temperature drops to 150 degrees F.

(BPI water) t - 2.3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months Electromatic relief valve isolated by operator after S.G.-B isolated to prevent leakage t - 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months RCS pressure increases to 2150 psi and electromatic relief valve openeJ t aa 3.25 hours2.893519e-4 days 0.00694 hours 4.133598e-5 weeks 9.5125e-6 months RC drain tank pressure spike of 5 asig t 3.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months RC drain tank pressure spike of 11 psi -

RCS pressure 1750; containment pressure increases from I to 3 psug t a 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Peak containment pressure of 4.5 psig t a 5 - 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months RCS pressure increased from 1250 psi to to 2100 psi

EVENT TIM valve to t - 7.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Operator opens electromatic relief of depressurize RCS to attempt initiation RHR at 400 psi psi t = 8 - 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months RCS pressure decreases to about Son Core Flood Tanks partially discharge containment 28 psig containment pressure spike, gal. of t a 10 hour1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months and stopped after 5O' sprays initiated operation)

NaOH injected (about 2 minutes of to repressurize t a 13.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months Electromatic relief valve closed RC pump RCS, collapse voids, and start to 230A psi RCS pressure increased from 650 psi t - 13.5 - 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months temperature t - 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months RC pump in Loop A started, hot leg and cold leg decreases to 560 degrees F, F.

temperature increases to 400 degrees indicating flow through steam generator Thereafter S/G "A" steaming to condenser Condensor vacuum re-established F.,

RCS cooled to about 280 degrees 1000 psi Now (4/4) High radiation in containment 4E0 All core thermocouples less than degrees F.

small Using pressurizer vent valve with makeup flow Slow cooldown RB pressure negative

I I Enclosure TE Bulletin No.79-05A Page 1 of 3 April 5, 1979 LIS TII OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Date Issued issued To Bul ietin Sutject No.

4/14/78 All Power Reactor 78-05 malfunctioning Of Facilities with an Circuit Breaker OL or CP Auxiliary Contact Mechanfim-General Model CR105X 5/31/78 All Power Reactor 78-06 Defective Cutler- Facilities with an Hammer, Type 1 laluys OL or CP With DC Coils

£/12/78 All Power Reactor 78-07 Protcticn afforded Facilities Uith an by Air-Line Respirators OL, all class E and F and Supplied-Air Hoods £et!iclt Feactors with an DL, all Fuel Cycle Facilities with an OL, and all priority 1 Material Licensees 6;12/78 All Power and 78-08 Radiation Levels frror Research Recctor FUel Element Transfer Facilities with a Tutes Fuel Element transfer tube and an OL.

6/14/78 ill. B6 ' Power 78-09 - BWR Drywell Leakage Reactor Facilities Paths Associated with with ar. C1. *r CP Inadequate Drywell Cicsti es pave!

6/27/78 All BWR 7E-10 Bergen-PaterLon Reactor Facilities VydTaulic Shock with ar O or C?

Suppressor Accumulator SFrlng Coils

. .I I; Enclosure XE Bulletin No.79-05A Page 2 of 3 April 5, 1979 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Date Issued Issued To Bulletin Subject No.

7/21/78 BWR Power Reactor 78-11 Examir.ation of Mark I Facilities for Containment Torus action: Peach Welds Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monti-cello and Vermont Yankee 9/29/78 All Power Reactor 78-12 Atypical Weld Material Facilities with an in Reactor Pressure OL or CP Vesrsel Welds 11/24/78 All Power Reactor Atypical Weld attr~ll 78-12A Facilities ulth at in Rezctor Pressure OL or CP vessel Welds 3119/79 All Power Reactor 78-12B Atypical Weld Material Facilities with an in Reactor Pr*ssure CiL or CP Versel Welds 10/27/78 All general and 78-13 Failures In Source Heads specific licensees of Kay-Ray, Inc., Gauges with the subject Models 7050, 7050B, 7051, ay-Ray, Inc.

7051B, 7060, 7060B, 7061 gauges and 7061B 12/19/78 All GE BWR facilities 78-14 Deterioration of Buna-N with an OL or C?

Components In ASCO Solenoids 2/8/79 All Power iriatcr 79-01 Environmental Qualifica- Facilities with an tion of Class IE Equipment 01. or CP A

k .

. 9 I I. i0l

s Enclosure IE biletin No. 79-OSA Page 3 of 3 April 5, 1979 LISTING OF 1E BULLETINS ISSUED IN LAST TWELVE MONTHS Date Issued Issued To Bulletin Subject No.

3/2/79 All Power P-eactor 79-02 Pipe Support Base Plate Facilities with ar.

Designs Using Corcrete OL or CP Expansion Archor Bolts All Power Reactor 79-03 Longitudinal Weld Defects 3/12/79 Facilities with an In ASHE SA-312 Type 304 OL or CP Stainless Steel Pipe Spools Manufactured By Youngstown Welding and Engineering Co.

3/30/79 All Power Reactor 79-04 Incorrect Weights fcr Facilities vith tr Swirg Check Valves 0L or CP manufactured by Velan Engineering Corporation 4/1/79 All B&W' Power 7S-05 Nuclcar Tncident Reactor Facilities at Three Mile Island. with an OL

I .( I t

. 1 *'Ik UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 APRIL 21, 1979.

IE Bulletin 79-05B NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Description of Circumstances:

Continued NRC evaluation of the nuclear incident at Three Mile Island i-==

Unit 2 has identified measures in addition to those discussed in IE Bulletin 79-05 and 79-05A which should be acted upon by licensees with reactors designed by B&W. As discussed in Item 4.c. of Actions to be taken by Licensees in IEB 79-05A, the preferred mode of core cooling following a transient or accident is to provide forced flow using reactor coolant pumps.

It appears that natural circulation was not successfully achieved upon securing the reactor coolant pumps during the first two hours of the Three Mile Island (TMI) No. 2 incident of March 28, 1979. Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of noncondensible gases, in the primary

. J. coolant system. To avoid this potential for interference with natural circulation, the operator should ensure that the primary system is ir subcooled, and remains subcooled, before any attempt is made to establish natural circulation. A.

[ .... .

Natural circulation in Babcock and Wilcox reactor systems is enhanced by E....

maintaining a relatively high water level on the secondary side of the A:::

once through steam generators (OTSG). It is also promoted by injection of auxiliary feedwater at the upper nozzles in the OTSGs. The integrated _ . .

Control System automatically sets the OTSG level setpoint to 50% on the :.-._

operating range when all reactor coolant pumps (RCP) are secured. However, A._

in unusual or abnormal situations, manual actions by the operator to increase steam generator level will enhance natural circulation capability .... _

in anticipation of a possible loss of operation of the reactor coolant pumps. ___

E-As stated previously, forced flow of primary coolant through the core is , _

preferred to natural circulation. .:_

c Other means of reducing the possibility of void formation in the reactor

=

coolant system are: ___

=

A. Minimize the operation of the Power Operated Relief Valve (PORV) on the pressurizer and thereby reduce the possibility of-pressure _

reduction by a blowdown through a PORV that was stuck open.

A__

. .- - , . _ ........... ......... _. En

- 1E Bulletin 79-05B April 21, 1979 Page 2 of 4 B. Reduce the energy input to the reactor coolant system by a prompt reactor trip during transients that result in primary system pressure increases.

This bulletin addresses, among other things, the means to achieve these objectives.

Actions To Be Taken by Licensees:

For all Babcock and Wilcox pressurized water reactor facilities with an operating license: (Underlined sentences are modifications to, and supersede, IEB-79-05A).

L

1. Develop procedures and train operation personnel on methods of establishing and maintaining natural circulation. The procedures and training must include means of monitoring heat removal efficiency by available plant instrumentation. The procedures must also contain a method of assuring that the primary coolant system is subcooled by at least 500F before natural circulation is initiated.

In the event that these instructions incorporate anticipatory fillina of the OTSG prior to securing the reactor coolant pumps, a detailed analysis should be donie to provide guidance as to the expected system response. The instructions should include the following precautions:

a. maintain pressurizer level sufficient to prevent loss of level indication in the pressurizer;

b. assure availability of adequate capacity of pressurizer heaters, for pressure control and maintain primary system pressure to satisfy the subcooling criterion for natural circulation;

c. maintain pressure - temperature envelope within Appendix G limits for vessel integrity.

Procedures and training shall also be provided to maintain core coolinga in the event both main feedwater and auxiliary feedwater are lost while in the natural circulation core cooling mode.

2. Modify the actions required in Item 4a and 4b of IE Bulletin 79-05A to take into account vessel integrity considerations.

"4. Review the action directed by the operating procedures and training instructions to ensure that:

a. Operators do not override automatic actions of engineered safety features, unless continued operation of engineered

.IEBulletin 79-05B April 21, 1979 Page 3 of 4 safety features will result in unsafe plant conditions. For example, if continued operation of en ineered safety features would threaten reactor vessel integrity then the HPI should be -

secured (as noted In b(2) below).

b. Operating procedures currently, or are revised to,. specify thatE if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:

(1) Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The HPI system has been inoperation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the exist4ng RCS pressure. If50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated. The degree of subcoollng beyond 50 degrees F and the length of time HPI is in operation shall be limited by-the pressure/

temperature considerations for the vessel integrity."

- 3. Following detailed analysis, describe the modifications to design and procedures which you have implemented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during anticipated transients. This analysis shall include consideration of a modification of the high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of the PORV for the spectrum of anticipated transients discussed by B&W in Enclosure 1. Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients.

4. Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant system. These transients include:

a. loss of main feedwater

b. turbine trip

c. Main Steam Isolation Valve closure

d. Loss of offsite power

e. Low OTSG level-,

f. low pressurizer level.

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4 a.

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IE Bulletin 79-05B April 21, 1979 C::=-.,

Page 4 of 4 FEEF t==

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5. Provide for NRC approval a design review and schedule for implementation of a safety grade automatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level.

6. The actions required in item 12 of IE Bulletin 79-05A are modified as follows:

Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with NRC.

7. Propose changes, as required, to those technical specifications which EHH-must be modified as a result of your implementing the above items. ECHE Response schedule for B&W designed facilities:

a. For Items 1, 2, 4 and 6, all facilities with an operating license respond within 14 days of receipt of this Bulletin.

t=.e

b. For Item 3, all facilities currently operating, respond within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . All facilities with an operating license, not currently operating, respond before resuming operation. =

c. For Items 5 and 7, all facilities with an operating license respond W in 30 days. !:. . '

Reports should be submitted to the Director of the appropriate NRC Regional

=.

Office and a copy should be forwarded to the NRC Office of Inspection and _

Enforcement, Division of Reactor Operations Inspection, Washington, D. C. _

20555. :.::::

For all other power reactors with an operating license or construction c permit, this Bulletin is for information purposes and no written response . _.

is required. _,

Approved by GAO, B180225 IR0072); clearance expires 7/31/80. Approval __-

was given under a blanket clearance specifically for identified generic -

problems. __

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EXTRACT OF B&W COK4UNICATION - RECEIVED BY NRC Enclosure I th[ROWCTION 4/20179 Page 1 of 4

,2K .OHYTI1UIllG REVIEI OF THE.SEQUENCE OF EVENTS LEADItnG TO T1E WICIDENT AT

__2 O.N t'iRC 28, 1979 SHOWS TMAT ACTIOtN CANl 8E TAKENt TO PROVIDE ASSURATimE r~ THE PILOT-OPERATED RELIEF VALVE (PORY) MiVUTED OI THE PRESSURIZER OF BEm PLAtTS 1ILL k'QT BE ACTUATED BY ANTICIPATED TRANtSIENTS WjICHfHAVE OCCURRED OR HAVE A SIGNIFICANT PROBABILITY OF OCCURRING III THESE PLANTS. T1IlS ACTION Itusr Tr DEGPAE THE SAFETY OF THE AFFECTED PLATS WI1t RESPECT TO THEIR RESPONSE TO OWLt,UPSET OR ACCIDEET CONDITONS NOR LEAD TO UNREVIEVED SAFETY COUCERNS.

THE AW'TICIPATED TRANSSIEITS OF CORCERN ARE:

i lSOS EXtERrLA ELECTRICAL LOAD

2. TURBIER TRIP

3. LOSS OF MAIN FEEIMATER

4. LOSS OF CONMDESER VACJUU.*

S. IlhDVERTENT CLOSURE OF rAIN STEAM ISOLATION V.ALVES (ff5IV).

A tWi3ER OF ALTERflATIVES WERE CONSIDERED IN DEVELOPING THE ACTIONS PROPOSED BELO'S. INCLUDING:

I. FESTRICTIM RACTOR POWER TO A VALUE WICH1C1 WOULD ASSURE NO ACTUATION OF THE PORP. THE. REACTOR PRGTECTION SYSTEM, DESIGN PRESSURE AND PORY SET-POINTS RE1UINED AT THEIR CURREIJT VALUES.

L_ LO.1ERING THE HrGH PRESSURE REACTOR TRIP SETPOINT TO A VALUE WHIC WOIULD fSSURE NO ACTUATION OF THE PORY. THE DESIGMl PRESSURE OF THE REACTOR AND

\.J>HE SETPOINfT FOR PORY ACTUATION REMAINED AT THEIR CURRENiT VALUES.

.. LO rEERING T HIGI PRESSURE REACTOR TRIP SETPOINT NU rADJUSTIrNG THE OPEPATING PRESSURE (ArND TEIPERATURE) OF THE REACTOR TO ASSURE ttQ PORV ACTUATIM ARMQ TO PROVIDE ADEQUAT7E MARGIN TO ACCOMIODATE VARIATIONS Itt OPERATINIG PRESSURE. THE SETFOINT FOR PORY ACTUATION RErAlMiED AT ITS CJRREUT VALUE. THIS ALTERNATIVE WOULD REDUCE RlET ELECTRICAL OUTPUT.

4. MADIfSTM THE HIGH PRESSURE TRIP AM THE PORY SETPOINITS TO ASSURE tU POV ACTUATION FOR THE CLASS OF AITICIPATED EVENTS OF CON:CERN. THE OESIGn PRESSURE OF THE REACTOR REKAIHED AT ITS CURRENT VALUE.

All ATIALYSIS OF THE I1hACT OF THEtE VARIOUS ALTERMATIVES AND THEIR CONITRIBUTIONI TO ASSURING THAT THE PORY WILL MT ACTUATE FOR THE CLASS OF ANTICIPATED TRANSIENTS DF COI;CERN? HAS BEENI CO"LETED. THE RESULTS SHOW THAT:

LO'riERIS TE HIGH PRESSURE REACTOR TRIP SETPOItl FRON 2335 PSIG TO 2330 PSIG PAiISI3G THE SETPOI1IT FOR TnE PILOT OPERATED RELIEF VALVE PFW2 2255 PSIG TO 2450 PSIG f 1HE REQUI.EO ASSURANCE. THIS ACTION HAS THE FURTIIER ADVANTAGES OF:

Enclosure 1 I . 1. , ' '.

EXTRACT OF B&W COMMUNICATION - RECEIVED BY NRC Page 2 of 4 -

4/20/79

t. REtCIt'9 TiE PROBABIL17Y OF PORY AND ASHE CODE PRESSURIZER SAFETY VALVE ACTLLMTIO FOR OTHER IflCREASIt PRESSURE TRARSIENrS.

2 0FRSERVIHG PRESSURE RELIEF CAPACITY FOR ALL HIGI PRESSURE TRAISiEWTS.

Q K FRINATiN'S THE OOSSIBILITY OF ItiRODUCING UNlREVIEWED SAFETY COtlEPERNS.

4. REDUCIG THE T11M AT WICH T14E STEAM SYSTEM HEAT S111K WOULD BE LOST Ill THE EVENT EKERGENCY FEED4ATER FLOW WERE DELAYED.

h stMtR OF TilE IMPACT OF THE PkPojED SMTPOINT CHAINGES ON ALL ANTICIPATED

TRAPISIErTS IS GIVEff II TABLE 1.

SUM PLA4TS ARE CURRENTLY CAPABLE OF RtfBACK TO 15% OF FULL POWER UPON LOSS OF LOAD OR TRIP OF THE TURBINE. THIS CAPABILITY FEQUIRES ACTLLATIOU OF T11E PILOT-OPERATED RELIEF VALVES. THE CAPABILITY INCREASES THE ELAOILITY OF POWER SUPPLY TO THE SYSTE BY RETUMMIG .THE UNITS TO POWER GElIERATIOlI MORE QUICKLY AFTER THESE TRA7SIERTS. THE ACTION PROPOSED ABOVE WILL REQUIRE THAT THE AEACTOR BE TRIPPED FOR THESE EVEfTS; NOTE:

The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typified by the attached figure 1. which was developed by.

B&W for a loss of feedwater transient.

. N.-_j.-I0 I

7 x.

I r~3Lr.

Page 3 of 4

S~tSt"RY OF PROTECTIO1. AGAINS5T PORY ACTUAT Uil~

PROVIDWE BY PRMSED SETPOINT CHANGES FOR ALL

-_ F.'I;LrAILU IKNslntDI E--*XTRACT OF B&l CQaUN'CATION - RECEIVED BY NRC 4(20/79.

i.. I;TICIPATED TRi SIENIrS VIICH HAVE OCCURREd AT B&1 PLAUTS AIWO M1IC11 WOULD tIVATE PORV AT THE CUPREKr SETPOINT (2255 PSIG):

LyORLL A. TUMIKE TRIP LOSS OF EXTELAL ELECTRICAL LOA C. Loss OF iMNI FEEMATER

v. LOSS OF CWNDE1NSER VACUY1h

E. 1hD VERTE11T CLOSURE OF t;IV A11TICIPATED TRA.'SIEfRS WHICH HAVE OCCURRED AT 8&W PLANTS AviO wl)1CH IMULD KOPPLALLY ACTUATE PORV AT DIE PROPOSED SEPOIWf (Z450 PSIG):

. sTICIPATED TMNSIEIWS MfICH RAVE 14OT OCCURREO AT BMW PLANTS (LOW P AO3SILIT EVENTS) AflD VIOI 1A.0 KORMLLY ACTUATE PORY AT THE CUAREffr SETPO IT (2255 iSIG):

A. SO? CO-HTRGL f00 GROUP WITHDRAWALS (MODERATE TO HIGh REACTIVITY YOT7H GROLJS GOT OTHERWISE PROTECTED BY HIGH FLUX TRIP).

B. DsRATOR DILUTIONI.

A. A.'V'tCIPATED TRAh6SIEfiTS V-HICH HAVE ItOT OCCURRED AT B&W PLrjuTS (LOW PROBAl13IlITY

- EVEtITS) AUD dt.iCHi MOJLD ACTUATE THE PORV AT THiE PROPOSED slPOINT (2450 PSIG):

SOE CO1rrROL P.00 GROUP IIMTHDRMALS (HIGHi REACTIVITY "OflrTi rIor

I OTHERIASE PROTECTED BY HIGH FLUX TRIP).

... .. . .. ~. . ........ .........

Enclosure I

. ' I .. Page 4 of 4

.EXTRACT OF B&W COMMUNICATION - RECEIVED BY NRC R I I

I I iI

- -; . . . 4 '. . . . II I . . .0 . .

- . o- . . . . .

.. .- 4. - __ . . . . . .

0 ... .. .

. . . I ... . '.. rwo"09 it. , --,

11 j. = . . .

t=.: -.1 -:-.-

. -. I :.--.-:'4: :" -:1

- 7.: - I : : '!

.:.. F':.- 1.- .: I-I . i. Piz 6 Ms wa IC',"I E

A-SIC L_ - . . _... # .. . i . . _... . A... I.. . i - .'r-o "SS _'

i :: Ik . _-_ - ._- -- - r A10'r '_ '

. . - - r . ... . ,

-A - - - I -- .. . - . .

- Z - d'L --

1w - - - - o

,--it a a - . .

..-. . - - . . - . . & . : . a . . . . . I.. . .

.. I 1W I

.6 ......

_f

. I r.- --1 :.

I E.'

6 . . .

I

--- i ' "I =--II .. m. - '1' - .- 1'--' . i I : I : :.7 ' . I....:

I

.. i . . !

'f A 77-,

7.

7-77.17.

3

':San; -I

.=t:

Orr 7 :1 r - . . . .

- :..-1: --. 1: : - ( :::.: : . I

_ --- -T.

I-- -. !J I .

a * . .

_ :_.::4 A--Z.... I .J:

t . .. :'. . I 0 . . .

piz es ep4r I F.--. : -- ,; ; 'S : .:

7

____ I

. .- =-

. .. 1.

I. . . .

I i . .4 F F :: ;'? 0,

.. -A A . L . .. - I -

I

___4 : I::;: 1: -'Rj

- Ea ?00 -1 . : -

0 A..

0 24 00 77-7 I- i 7-. *W" RCS P*q'C'r-Selm r "I . . 0& I - . . . .

77'- -

=...- - --- 1 .. - . 67777t-

.:-1: I' . 3 . 0 ..I ---I Peak pressurizer pressure -:: at a1::::

.__1o.

?-Z initial pressures.

r=

for a loss of feedwater transient for expected conditions and various function of RCS pressure trip setpoint

- ..Figure1:.. 1

ML031210600

Text

Enclosures:

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