ML18142B977

R. E. Ginna - 02/24/1977 Response to Request for Additional Information Reactor Vessel Overpressurization

ML18142B977
Person / Time
Site:	Ginna
Issue date:	02/24/1977
From:	White L Rochester Gas & Electric Corp
To:	Schwencer A Office of Nuclear Reactor Regulation
References
Download: ML18142B977 (59)

Text

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FILE NUMBER NRC DISTRIBOTION For. PART 50 DOCKET MATERIAL TO: A. SCHWENCER FROM: ROCHESTER GAS 6r ELEC. CORP. DATI OF DOCUMLNT ROCHESTER, N.Y. 22477 L.D.fAIHITEFJR. DATE RECEIVED 3 1 77 l3LETTE B 0 NOTOB IZE D PROP INPUT FORM NUtvlGER OF COPIES RECEIVED CRO B I G IN A L DVNCLASSIFIED QCOPY DESCRIPTION ENCI.OSV RE LTR. RE. OUR 1/10/77 LTR...TRANS THE FOLLOWIH RESPONSE TO STAFF POSITIONS AND ADDITIONAL INFORMATION REQUESTS REGARDING THE REACTOR VESSEL OVERPRESSURIZATION.....

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<> ui>ijjz>i,iiiwziii i znidiN'iiiiiiiiiiii ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, M.Y. 14649 LEON O. WHITE, JR, TELEPHONE VICE PRESIDENT AREA COOC 714 546-2700 h

February 24, 1977 E

4 pg Director of Nuclear Reactor Regulat 18PPp MAR l>SPt Attention: Mr. A. Schwencer, Chief bOlg bR MERE 5UCKAR REGNA'885 Operating Reactors Branc lO ~isseN 7 U.S. Nuclear Regulatory Commission M!Secuog Washington, D.C. 20555

Dear Mr. Schwencer:

In your letter of January 10, 1977 you directed us to select one of two options for implementation of hardware improvements to meet your objective of improved overpres'sure 'protection in all operating PWR facilities by the end of 1977. These hardware 'improvements are to be in addition to the procedural and administrative measures which we have already instituted and which you concluded will help prevent any future pressure transients.

We are designing and will procure and install components for an overpressure protection system which will meet the design criter'ia out-lined in Attachment 2 .to your January 10, 1977 letter, to the extent practical given the existing equipment configuration of R. E. Ginna.

The system will be based on the assumption that a single pressurizer power operated relief valve will mitigate the consequences of over-pressure transients not caused by'nadvertently discharging an accumula-tor. To the extent practical, we will install redundant overpressure protection components which are seismically qualified, which meet IEEE-279 criteria and which can be tested on a schedule consistent with the frequency of use for overpressure protection. The objective will be to provide a system which is not vulnerable to an event which both causes a pressure transient and causes a failure of equipment. needed to terminate the transient.

No shutdown has been scheduled for Ginna during other than the 1977 Spring refueling outage which is the to coming year commence in April. Components for the overpressure protection system are not avail-able for that shutdown. Because the overpressure protection system will not be used while the plant is operating and is of value only during shutdowns, we do not plan to schedule a special outage to install this equipment.

8073

P ROCHESTER GAS AND ELECTR ORP. SHEET NO.

oA'TE February 24, 1977 To Mr. A. Schwencer, Chief Therefore we plan to install the long term hardware improvements described above during the first scheduled shutdown after December 31, 1977, provided the system components have been delivered to us and you have concurred with our response to your letter of February 14, 1977.

We will work expeditiously to have the components available by the end of this year. Should an unscheduled outage of sufficient duration to install the system occur before that shutdown but after the system com-ponents have been delivered, we will install the system at that time.

Enclosed is the additional information which you requested in to your January 10, 1977 letter.

Sincerely yours,

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RESPONSE TO STAFF POSITIONS AND ADDITIONAL IiiFOK1RTION REQUESTS (Attachment 1 to A. Schwencer letter dated January 10, 1977)

REACTOR VESSEL OVERPRESSURIZATION R. E. GINNA UNIT NO. 1 DOCKET iNO. 50-244 February 24, 1977

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R. E. GINNA NUCLEAR PONER PIANT

- 'QUESTIO 7 1". The staff cons'iders it. essential that all plant operators (i..e., reactor operators, ecuipment operators, Instrument & Control personnel) be made aware of the details of the pr'essure tran-sients which have taken place at all PNR facili-ties. POSITIO".}: Pormal discussions should be held with t'e operator to review the causes of past pressure transients that, have occurred at other operating "PNR. facilities'our discussions should,.include the plant conditions at the time, the mitigating action that could have been or was taken, and the preventive measures that could have been taken to avoid the event, and the steps taken to prevent, similar, further occurrences.

Plant similarities and distinctions should be along with how these relate to plant start- iden-'ified up, shutdown,, and testing operations. Nith regard to this position, you are requested to provide the following information:

a'. If you have not already completed the required formal discussion, when will you do so?.

b.. -How will the discussions be held?

c. Of the past PNR Appendix G violations that have occurred at PNR facilities and which are des-cribed in License Event Reports, identify which are not credible in your plant due to eauipment differences. Provide.a desciiption of the dis-tinctions.

d. Describe, in detail, how you are reducing the o f the other remaining credible .;<'ikelihood events. Furnish schematics, diagrams or pro-cedural summaries necessary to support the effectiveness and reliability of these measures.

RESPONSE: l.a.

The formal discussions called for in the Staff Posi-tion were begun the week of P'ebruary 8, 1977 and .

are scheduled to be completed the week of .""larch 8, 1977.

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RESPONSE: 'l.b.

(Continued) . The discussions 'are being held during Training and will include all Operator'equali'fication licensed. operating personnel and Xnstrumentation and Control technicians. The discussions include similarities to 'and differences from the R. E.

Ginna Plant and empnasize the actions that could have been or were taken to.reduce the impact of the overpressurization event. Preventive measures which can be taken to prevent similar events are also discussed. Plant startup, shutdown, and testing conditions are related to the potential for over-pressurization events.

l.c.

Of the past PNR Appendix G violations that have occurred at PNR facilities, s' are considered to be incredible events at. R. E. Ginna because of equipment differences.

Events simila'r to those which occurred at Beaver Valley Unit No. 1 on February 24, 1976, Turkey Point Unit No. 3 on December 3, 1974, and Zion .Unit No. 2 cn September 18, 1975 were caused by"automatic iso-lation of the RHR system. There is no automatic-isolation of RHR at Ginna.

The event. at indian Point Unit'o. 2 on Nay 18, 1973 resulting from closure of certain air operated

.valves in the reactor coolant letdown system was caused by freezing of moisture in the air supply line. At Ginna the entire instrument air system is inside heated buildings and the air passes through air dryers before being piped, to the point of use.

The event at rojan on Julv 22, 1975 resulted from.

the PHR suction valve from the reactor coolant sos-'~

tern being closed and isolating system letdown whi'le the positive displacement charging pump iras operating.

At Ginna, isolating the RBR suction line from the reactor coolant system would not. isolate letdown. A flow path exists from, the reactor coolant system through

'the RHR discharge line to the RHR relief valve. This relief valve will limit system pressure to 600 psig and prevent exceeding technical specification limits whenever the reactor coolant system temperature is greater than 175'F.

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e RESPONSE: The event at Peach Bottom Unit No. 2, a boiling (Continued) w>ater reactor, on .~larch 6, 1974 is considered not to be applicable to Ginna, a.pressurized water reactor., because of the different reactor types and operating eauipment. Xt is possible for an system while holding operator to cool the primary constant system pressure although an alarm has a

been insta"led to alert the operator limits.

if he approaches Technical Specification lode The remaining events which remain credible, although improbable, at Ginna have been caused by inadvertent, letdown isolation, reactor coolant pump starts, in-advertent accumulator injection, safety injection tests or improper procedures or personnel error.

The likelihood of letdown isolation overpressurization events similar to those which occurred at indian Point, Unit. No. 2 on 2/17/72 and 4/6/72, Prairie Island Unit, No. 2 on 11/27/74, St. Lucie Unit No. 1 on 8/12/75, D. C. Cook on 4/14/76 and Beaver Valley Unit No. 1 on 3/6/76 is reduced by providing a second let-down path. The cooldown procedures require that the RHR valves to the primary system be opened while there is a steam bubble in the pressurizer. The procedures also reauire that the RIM system be connected to the letdown system which contains a 600 psig relief valve. The analysis in Appendix A to this attach-ment has shown that this valve and the piping.con-necting it to the RHR system are capable of reliev-ing the full output of two charging pumps (120 gpm) over the design temperature range of the RHR system. This valve and piping can relieve the output of three charging pumps (180 gpm) when the system tem-perature is below 200'F. Plant procedures control the use of the charging pumps within these guidelines while the RHR system is in operation.

The plant heatup procedures call for a steam bubble to be formed in the pressurizer prior to isolation of the RHR system. There is no automatic function wnich will isolate the RHR system upon rising primary system pressure.- Thus the relief valve will remain available.

The likelihood of overpressurization events caused by reactor coolant pump starts similar to those at indian Point Unit No. 2 on 3/8/72 and 1/23/74, Prairie Island Unit 'No. 1 on 10/31/73 and St. Lucie Unit No. 1 on 6/17/76 has been reduced by procedural control of the 1-3 g/24/77

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RESPONSE: reactor coolant pumps. Xf all the reactor coolant (Continued) pumps have been idle for more than 5 minutes and the reactor coolant temperature is greater than the

'charging anQ seal injection water temperature, a-

~ 4 "steam bubble must be formed in the pressurizer before starting a reactor coolant pump. ln addition, if the reactor coolant is being cooled by the RHR system and all the reactor coolant pumps have been idle so that a nonuniform temperature distr"'bution may have occurred in t¹ reactor coolant system, a steam bubble must, be formed 'n the pres urizer prior to starting a reactor coolant pump. Our.cooldown procedures call

. for the last, reactor coolant pump to be stopped only after the reactor coolant system has reached approxi-mately 150'F, thus.minimi.zing any temperature differen-tials which could occur be'tween system components. On startup, the operators norm'ally throttle back RHR cooling "

and allow the reactor temperature to rise.

Overpressurization events similar to those at Indian Point Unit No. 2 on 2/22/74 and Prairie Xsland Unit No. 1 on 1/16/74 resulting from inadvertent: discharge of an accumulator have, been reduced in likelihood by, deenergizing the accumulator isolation valve at all es when the reactor coolant system pressure is less than approximately 1000 psig. The cooldown procedures require that the discharge valves between the accumula-tors and the primary system be closed, at less than..1800 psig and that power be removed from these valves at approximately -1500 psig. The heatup procedures do not allow the accumulator isolation valves to be reenerg'ized and opened until primary system pressure is greate'r than 1000 psig.

An overpressurization event caused by an improperly aligned safety injection system during tests similar to that at Point Beach Unit No. 2 on 12/10/74 is un-likely at Ginna because of the procedural controls applied to periodic tests. The test procedures require that. whenever reactor coolant system pressure is less than 1600 psig during safety injection pump tests, the pump discharge, valves must. be closed and the

'initialed by the operator to signify that the procedure valves have been properly aligned.

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I RESPONSE: "The remaining six.overpressurization events 'whose

,(Continued)~:-'."causes are known resulted from=what appear to have been poor operating procedures, operator .improvised procedures or operator error. Ginna operations and tests procedures are reviewed by the Operations Engineer.,or Tests and Results Engineer respectively and are reviewed by:the Plant Operations Review Committee. Operators are not allowed to improvise or deviate from established procedures. Procedures may be revised on a temporary, basis only as de-tailed in Technical Specification 6;8.3. Operators are keenly aware of overpressurization events be-cause of recent increased emphasis and training.

The'ffectiveness and reliability of the"measures which have-been taken to reduce the. likelihood of =,

overpressurization events is demonstrated by the Ginna operating history.

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'xlUCLEAR POL'lER PLANT QUESTION '2: The majority of the reported pressure events have occurred while the plants were operating in a water solid conditicn. POSITION: The staff will require that operations during which..the plant is maintained in a water solid condition be minimized or if possi-ble eliminated. Those operations in which the plant is in a water solid condition must be fully justified.

Attachment 1 to your October 1S letter and your October 26, 1976 letter to the NRC discuss various requirements that minimize the operation of R. E.

Ginna while in a solid'-water condition, Based on the staff's position stated above, and the review of your letters, the following additional informa"ion is requested.

a. Your justification for n'ot partially draining and venting the pressure during cold shutdowns of less than seven days duration.

b. Describe the procedures, evolutions or situations

.that require the plant be maintained in a water solid condition.= Also provide reasons whv a nitrogen, air or steam bubble cannot, be main-tained in these situations.

c ~ Include sufficient background or 'supplementary information such as system diagrams, procedure summaries and descriptions. of equipment operation to justify your need for operating'he plant water solid.

RESPONSE: 2 oa ~

Draining and venting the pressurizer lengthens the amount. of time required to return the plant to opera-

-tion by approximately .8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The addi" ional. time is required to "burp" the pressurizer and to return the water chemistry to the specified limits. Cold shutdowns of short duration are occasionally required to repair or a'djust equipment., Occasionally unolanned outages occur which may require investigation to de-termine the extent- of the difficulties. Seven days is sufficient time to locate and correct, minor pro-blems or to determine that draining and venting the primary system is'equired.. Draining and venting the primary system generates additional liquids and gasses which'must be processed. The gasses are eventually discharged. Thus, a limit of seven days will minimize the length of time for water solid plant operation and also minimize the quantity of radioactive effluent.

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RESPONSE: 2.be (Continued) The plant must be operated in =a water solid condition during filling and'enting of the reactor coolant system,'hydrostatic pressure testin'g of the reactor coolant system boundary and during plant heatup in-cluding reactor coolant..-pump .operation prior to bringing the reactor coolant system within the chem-istry specifications..

Nitrogen, air or a steam .bubble is not maintained in the pressurizer during these conditions for the following reasons.'During filling and venting the pressurizer vent is used to release trapped air from the system. During the hydro test visual inspections for leaks are required on valves and piping on,"op of the pressurizer. Technical Specifications. pro-hibit drawing a bubble during plant heatup until reac-tor coolant system chemistrv specifications have bein met.'ir or n'trogen'n the pressurizer at higher temperatures is prohibited and therefore'ust be ex-cluded while bringing the 'coolant chemistry. within limits.

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The need for water. solid operation during filling and venting is evident.

The. reason for hydrostatically testing with water so-lid conditions is clear, however, it should be noted that this procedure is carried out, at temperatures high enough so that the pressurizer safety valves pro-tect the svstem from. exceeding echnical Specification limits.

Reactor-coolant pump operation whi.le water solid is required to circulate the coolant to aid in meeting the chemistry specifications. As noted in the response,.

to question 1, however, our cooldown procedure ca3.1s, ':.':;

for a reactor coolant pump to be run until the syst:"em ""

temperature reaches approximately, 150'P. This mini-mizes any temperature differences which may exist be-tween'he reactor and'team generators and minimizes

. the likelihood of an oyerpressurization event.

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QUESTION 3: The inadvertent operation of SIS components during cold, shutdown conditons has been responsible for a -major'ortion of the overpressure incidents.

POSITION: Based on the licensee submittals, the recent November 3-5, 1976 meetings, and discussions with NSSS vendors, the staff will require tne de-energ'ing of SIS pumps and closure of SI header/

discharge v'alves during cold shutdown operations.

Those situations during which this position cannot be met must. be described and be fully justified.

Your letters to the NRC dated October 15 and 26, 1976, discuss the relignments you make to the SXS during plant cooldowns =and heatups. The staff has reviewed .these actions; and requests the following additional information:

a. A schematic diagram of the SIS showing the flow-paths into the RCS.

b. The head-flow characteristics of each of the SIS pump s ~

c. Justification for placing the high head SI pump control switch'n "pull-stop" and not de-ener-gizing the pump.

d. Identify on the schematic diagrams the pumps and the valves to be closed, de-energized, or disabled, (e.g., "pull-stop").

e. Your time schedule for inplemen'ting the proce-,

.dural'r administrative changes that require the alignment shown on the above diagram.

Xndicate all circumstances for which these SIS pumps and valves may not be isolated, de-energized

. and/or disabled. For those situations, describe the manner in which SIS injection would be pre-

.vented.

g. The locat'on of the high head SI pump control switch, and all other places from which the pump may controlled (e.g., control center, local-

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~~lotor ly, etc.).

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,QUES ION 3: h. remaining SIS component (Continued) , power supply. breakers, and the places, from-which they can be controlled.

i. Describe=the position 'indication and status if signals which would be los" the components were (1) de-energized. or (2) disabled by placing control'sw'tch 'in "pull-stop".

j. De'scribe in detail, the administrative proce-dures which ~vill be used to assure proper equip-

.-"ment 'alignment 'and the supervisory personnel

'esponsible for 'maintaining control.

k. Provide, the reasons for. not closing and removing power- from the accumulator YiOVs at the same RCS pressure (as noted in you'r October 26'sub-

'mit tal) .

Describe the impact on overall plant operations.

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if you routinely lowered accumulator nitrogen pressure when in a cold shutdown condition.

,RESPONSE: 3. a' schematic of the 'safety injection system .is given in Figure 6.2-1 of the R. E. Ginna FSAR.

~ 3.b, The head-flow characteristics of the.'safety. injection pumps are given in Figure 6.2.-13 of the R. E. Ginna FSAR.

3ec these pumps .in "pull-stop" will prevent

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'lacing inadvertent actuation of the pumps and yet leave them available from the control room should they be needed. This method of operation provides greater core'protection and assurance of sare shutdown than does pulling 'the pump.breakers.

3~d~

. During plant cooldown from hot .shutdown to cold shut-.

dorm the accumulator discharge valves,865 and 841, are closed and the breakers are pulled. Safety in-jection pump discharge valves to tne cold legs, 878B and 878D, are closed and the safety injection pumps are put in th "pull-stop" position. Techn'ical Specifications require that safety injection pump discharge valves to the hot legs, 878A and 878C, be closed-.'and .deenergized for plant operation. They are not repositioned during=cooldown.

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RESPONSE:" 3.e.

(Continued) The alignments described in:the response to question

"..-.. 3.d'.are .those required by .existing procedures.

3.f.

A high head safety injection pump is used during fill to the accumulators to'heir'.normal cold'hutdown operating level. The procedure for filling the accumu-

.lators directs that, the pump discharge valves 878A, .

878B; 878C,- and'878D=be closed prior'to starting a..

pump 1 >>

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A refueling shutdown surveillance procedure covering di'esel generator. loading is conducted with the safety injection pumps'nergized and.'unning'in the re'circula-tion mode. Prior.to the- test the 878A, 878B, 878C and 878D valves are closed -or verified closed and the D.C.

control circuit fuses pulled'. The A.C. power is also removed'rom the valves.

3>>g>>

The high* head safety injection pump control switches are located in the control room. There are no loc'al, control switches outside the control room, however, the pump breakers at the 480 volt busses may be closed manually with the aid of a special handle.

3. h.

The power supply breakers for the valves discussed in 3.d above are located i'n the auxiliary. building in .

motor control centers 1C and 1D. They are'ontr'oiled only from the main control board.

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Valves that are deenergized by pulling breakers retain all of their normal staf us light indication. This, "..-

status indication is provided the D.C. contxol circuitry. Light indication isby lost only when the D.C.

control power fuses. are- removed at the motor'ontrol center breaker panel.

'hen the safety injection pumps'are p'laced in "pull-stop" the switch indicating light goes out and a ".safe-guard equipment locked off" annunciator is alarmed.

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RESPONSE: , 3.j.

(Continued). A'dministrative. procedure A-30-.4-,- Plant Procedure

.'- Adherence governs the use of the system

.* .: 'l'ignment Rendu'irements, procedures. Deviations'rom established

-procedures are permitted only'by the approval autnority outlined in Technical Specification 6.8.3. 'roper valve alignment is documented by the completed procedures. The supervisory per-

• 4 sonnel responsible for maintaining control are

.the Shift Foreman, Oper'at:ions Supervisor, and" Operations Engineer.

3.k.

There is- no reason except for procedure flow and continu'ity. .The valves are closed and power removed at pressures above the discharge pressures of the accumulator s.

3'.l.

The impact on overall plant operations of routinely.

lowering. the accumulator nitrogen pressure to one atmosphere during cold shutdown would be the loss two days to refill the accumulators to-of'pproximately operating pressure. This operation would cost several thousand dollars not including purchase of replace-ment, power during those times when it critical path operation. The operationwould be a involves an additional operator full time and a fitter part time to restore the accumulator nitrogen pressure'.

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QUESTION 4: The staff has noted that several Appendix G viola-tions have occurred during component or system tests while in cold and shutdown conditions. In, this.re-gard, please address the following questions.

a. 7fna+ components or svstems that could cause over-pressure transients, are r'outinely tested while in a cold shutdown condition'P

'..-What extra measures are taken to prevent an over-pressure event during these tests?

i RESPONSE: 4.a.

.The Diesel. Generator Load and Safeguard Sequence Test is conducted at cold shutdown. During this test the high head safety injection pumps are operated on re-.:<

circulation- flow. The safety injection pump discharge headers are -pressurized to approximately 1500 psig up to the hot, and cold leg injection valves 878A, 878B, 878C and 878D.

4.b.

To prevent an overpressurization event during this test the hot'and co'ld leg injection valves are pro-cedurally maintained closed with both D.C. control and A.C. motive power.'off.

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R. H. GXNNA NUCLEAR PONFR PLEAT QUESTXON 5: The staff believes that a high pressure alarm used during low RCS temperature operations is an effec-tive means to attract the operators's attention to a transient in progress. POSXTXON: The staff is requiring that if it is not presently installed, must be as soon as possible.

it Based on the staff's position stated above, and the review of your October 15 and 26, 1976 letters, the items listed below should be addressed:

a. The alarm setpoint, mode. of annunciation and sen-sors used.

b. A synopsis of the system modifications. that were necessary to furnish the alarm.

c. Your means to assure the alarm's availability during all water-solid operations, and to its downtime for all other cold shutdown mini-'ize conditions.

RES ONS> 5.a, Two setpoints are incorporated into the alarm. One is variable and follows the Technical Specification limit.. The other alarms at a given differential pressure, determined by the operator, below the Technical -Specification limit.- - Both setpoints alarm

'the bell and 1'ight on the computer'typewriter. The sensors'sed to generate the alarms and the Techni-cal Specification setpoint are the primary coolant loop wide range pressure transmitter and two cold leg.temperature sensors.

5.b.

The alarm was installed by the addition of computer

.software. No na dware changes were required.

5. c.

Xt is normal 'plant practice not to perform computer maintenance during changes of plant operating modes.

Computer maintenance is 'normally performed when tne plant is not in a water solid conditon.

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R. E . GIHNA NUCLEAR"POM'.R PLANT QUESTlON 6: The RHR (or SCS) is normally con>>ected to the PCS and operating when the plant is .in a cold shut-.

down condit'n. The inadvertent isolation of the RHR system while water solid has caused a number of overpressure transients, .and the RHR safety valve has actaallv terminated others: The RHR (or SCS) therefoxe plays an important part in the initiat'ion and possible mitigation of the Ph'R overpressuriza-tions. Accordingly, .we reauest the following addi-tional information:

a. RHP. (or SCS) design pressure.

b. A description of the system isolation valves and their arrangement (e.g., number and, configuration of valves installed, and pneumatic or motor oper-ated) .

c. interlocks, interlock setpoints, and alarms asso-ciated with each isolation valve.

d. Nominal'stroke time of isolation valves.

e. The setpoint and capacity of RHR (or SCS) relief and safety valves.

f..-All-pressure alarms; setpoints and -associated an-nunciation for the system.

RESPONSE: 6.a.

,The RHR..system design pressure is 600 psig.

6.b. \

The. RHR suction and discharge. valves connecting this syst: em to the primary coolant system are'hown on Figure 9 3-1'f the R. E. Ginna FSAR. The reactor coolant system suction supply to the RHR pumps is from the hot leg of loop A -through motor operated.

valves > O'V 700 and HOV 701 in series. The RHR pump

..discharge return to the loop B cold leg of the reac-tor coolant system is through two series motor operated.

valves, YOU 720 and MOV 721-.

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f RESPONSE:'.c, (Cont'inued) . Permissive interlocks required to open the four RHR system isolation valves are listed below.

"IfOV 700 (1) Reactor coolant system pressure.

must be less than 410 psig (2) RHP. suction valves ?<OV 850 A and HOV 850 B.from the containment

'sump must be closed.

NOV 701 (1) BHR suction. valves YiOV 850 A and llOV 850 3 from,the 'containment sump must be closed (2) the valve is operated by a key switch Y.OV 720 No interlocks exist but the valve is operated by a key switch i4OV 72 1 (1) Reactor coolant system pressure must be less than 410 psig Yo interlocks are associated with valve closure.

There are no automatic functions which close the valves and no alarms generated by the valves.

6.d.

The nominal stroke time of the RHR isolation valves is 3 minutes.

6.e.

The setpoint of the RHR relief valve is 600 psig.

The design capacity is 70,000 lbs/hour. A more de-tailed analysis of the relief valve is .presented in Appendix A.

6.f.

The pressure alarms, setpoints and associated annun-ciation of the RHR systen are a high pressure alarm at 550 psig and a'eac'tor- coolant system interlock pressure alarm at 410 psig,.

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R. E. GINNA NUCLEAR POL'KR PLANT QUESTION 7: Reactor coolant system heatups, resulting from im-proper operation of the reactor coolant pump (RCP) while in a cold, shutdown and water solid condition, have been responsible for approximately 15$ of the RCS overpressurization events. POSI ION: 4'e will require that all licensees include adequate provi-sions to prevent RCP starts while in a water solid condition unless absolutely necessary. 1n those cases where the PCP starts cannot be avoided, the licensee should take appropriate steps to cetermine and minimize the RCP temperature profile.

Based on the position stated above, please provide

.the following information:.

a ~ Describe the normal operating conditions during which the RCS is maintained wa'ter solid with all RCP's stopped (e.g.,

cooldown).

fill and vent, pressurizer

b. For each of the above procedures, justify your inability to establish a N2, air or steam bubble in the pressurizer prior to the start of the first RCP.

C Nhat are the limits associated with system temper-first RCP

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atures before the can be started in a solid RCS?

d.'pecify the instruments utilized to determine the RCS temperature profile.

e. Prov.'de the necessary schematics and procedural descriptions that show what your actions would be

.to bring the RCS to an isothermal condition.

Summarize any other measures you take to reduce possible RCS pressure spikes during RCP starts, (e..g , open all,letdawn orifice isolation valves, stop makeup flow, etc.)

RESPONSE: ~ 7~a ~

The reactor coolant system is maintained water solid with all reac"or coolan+ pumps stopped during (1) normal cold shutdown below approximately 150 F when the reactor coolant system is not to be opened, (2) fill heatup.

and vent operations and'3) the start of system 7-1 2/24/77

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RESPONSE: 7.b.

(Continued) A steam bubble can be formed in the pressurizer prior to the start of a reactor coolant pump pro-vided that the reactor coolant chemistry is within technical specification limits. If the chemistry is not v7ithin the eguired limits a reactor coolant pump must be started p"'ior to drawing a pressurizer bub'e in'rder to mix tne 'coolant and'the chemicals which are aQQed to restore proper system chemistry.

Air or N2 bubbles are not used because of the problems they cause in meeting chemistry recuirements. Thus, following those system operations listed above in 7.a., a steam bubble will be formed in the pressurizer before starting a reactor coolant pump provided that the reactor coolant system chemistry meets T chnical Specification limits.

7+co There are no formal limits on reactor coolant system temperatures prior to tne start of the first reactor coolant pump in a water solid system. However, the operating methods discussed in l.d. minimize tem-perature differentials between system components.

7.d.

Instruments used to determine the reactor coolant system temperature profile include pump seal injec-tion temperatu e, charging inlet flow temperature, hot and cold leg temperatures and RHR heat exchanger inlet and outlet temperatures.

.7+em As discussed in l.d.,

the steam generators the reactor coolant system and are= cooled to approximately 150'P before the last coolant pump is stoppeQ. The reactor temperature is normally allowed to rise prior to restarting the first coolant pump.

If the charging and seal injection water temperature is less than the reactor coolant temperature and both reactor coolant pumps have been stopped for more than five minutes,. a steam bubble is drawn. in the pre ssuriz er before starting the first. reactor coolant pump. This methoQ obviates the need for an isothermal condition.

7.f Before starting a reactor coolant pmnp in a water solid'oridition, the normal 'isolation valve and all 3 orifice isolation valvesletdo~m are opened. The operator takes manual control of the letdown pressure control valve. The procedures also caution the operator that.

will aid in re'ducing pressure spikes to reduce "charging it pump speed and stop the running RHR pump.

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APPENDXX A As can be seen'n Figure 9.3-1 of the FSAR, the RHBS is connected to the loop A hot leg on the suction side and the loop B cold leg on the discharge side. The design pressure and temperature of the RHRS is 600 psig and 400'F.

The design basis witn regard to overpressure protection for Ginna Station's RHRS is to prevent opening, of the, RHRS isolation valves when Reactor Coolant System (RCS) pressure exceeds 450 psig and to provide relief capacity sufficient to accommodate thermal expansion of water in the RHRS and/or leakage past the system isolation valves.

Nestinghouse Electric Corporation has performed an analysis of incidents which might. lead to overpressurization of the RHRS. Three events were considered in the analysis:

(1) The letdown line is isolated from the RHRS solid, the RHRS is in operation, and the charg-ing pumps are running.

(2) During cooldown utilizing the RHBS, one cooling train suffers a failure at a time when the heat generation rate exceeds the heat removal capability of the remaining cooling train.

(3) Pressurizer heaters are energized while the RHRS is in operation.

The results of the analysis are presented below.

Event, No. 1 The first event considered in the RHRS relief analysis is

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the isolation of letdown flow with charging pumps running and with th'e plant in a solid water condition and the RHRS in operation. For this analysis it was assumed that, once open, RV203 shuts at 5% blowdown, corresponding to a pressure at 570 psig for a setpoint. of 600 psig. Flow through RV203 at this pressure would be approximately 185 gpm with,the system below 200'F and 130 gpm with the system at 380'F,:the valve design point, based on letdown'emperature.

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The:analysis shows, the combined capacity of the 3/4 inch and. 2 inch branches connecting the RHRS'to CVCS (see Figure 1) and relieving through RV203 to be sufficient to pass the full output of two charging pumps (120 gpm) over the design= temperature'range of the RHRS (see Figure 2). Sufficient capacity to relieve the output of three pumps (180 gpm) when the system is below 200'F was also shown. PPnen momentary transient pressures due to relief valve

=.cycling are neglected, the RHRS .will remain below the maximum allowable p=essure of 660 psig (110-o of design pressure) during relief operation. This event can be terminated, of course,, by establishing letdown flow.

Plant procedures have been revised to prohibit use of three charging pumps while the RHRS is in operation.

Event No. 2 The" second event analyzed is that during a cooldowni utilizing the RHRS, one cooling train suffers' failure at a time when the heat generation rate exceeds the heat removal capability of the remaining cooling train.

The assumptions made for this analysis are:

a) RCS temperature and pressure are 350'F and 425 psig prior to the event.

b) The RHRS is brought into service, at 2 hours after shutdown (assumes a cooldown rate of 100'F/hr).

c) Immediately after initiating RHRS operation, necessary to isolate one RHR train (one RHRS pump it is and one RHRS heat exchanger).

d) For purpose of analysis,'no credit is taken for heat removal. via the steam generators once the RHRS is brought into service.

e) One RHRS train (one pump and one heat. exchanger) removes heat at, its design rate. This's conserva-tive because the high temperature difference 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after shutdown will result in more heat transfer.

f) Heat input to the reactor coolant system consists of decay heat (constant rate at' hours after shutdown),

thick metal heat, and heat from two running coolant pumps (3 l5~ each) .

g) RHRS relief valve opens at 600 psig and reaches a relief rate of 139. 2 gpm at 660 psig.

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. h) No letdown flow is assumed.

i) The pressurizer is water-solid prior to,the event.

The results of this. analysis are depicted in Figure -";. This analysis indicates a release of 90.2 gpm through the RHRS relief valve which with heat removal at the design rate of one RHR "rain (one pump and one heat. exchanger) would limit the coolant pressure to 638.8 psig.

It is anticipated thai. the relief valve may chatter during the course of this event. Figure 4 shows that the pressure quickly rises to a plateau and remains there. The constant heat source, described in assumption (f) causes a linear rise in average coolant temperature. This temperature and pressure response indicates that the coolant expansion rate is just matched by the water relief through the RHRS relief valve and the heat removal via the RHRS heat exchanger.

Only a small opening in the relief valve is required to limit pressure under these conditions. If more heat removal by the operational RHRS train is permitted (see assumption e), the relief valve may'emain closed.

Event No. 3 The thir'd event analyzed is energizing the pressurizer heaters while the RHRS is in operation. The assumptions made for this analysis are:

a) RCS temperature and pressure are 350'F and 425 psig prior to the event.

b) Only one RHRS train is in service (one RHRS pump and one RHRS heat exchanger).

c). No heat removal via the steam generators.

d) No heat removal,via the. RHRS.

e) The pressurizer heaters input heat at a constant rate

'of 800 kw until they automatically shut off when the pressurizer water volume drops below 85 ft3.

A letdown flow of 30 gpm is assumed in'ord'er to simu-.

-late the conditions under which. such an event is likely to occur heaters a e turned on during startup, and letdown flow is permitted in order to'raw a steam bubble in the presurizer. The expansion rate due to

.boiling in. the pressurizer is greater than that due to heating the pressurizer.

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g) Two reactor coolant pumps are= running and contributing 3 HN of beat/pump.

h) RHRS relief valve opens at. 600 psig and reaches a relief rate of 139.2 gpm at 660 psig. The relief rate is extrapolated linearly beyond 660 psig.

i) The pressurizer is water-solid. pr'r to the event.

j) No decay heat.

k) The operator response to the overpressure alarm on the RHRS, which is set at 550 psi, at ten minutes after the alarm is actuated and turns off the heaters.

As shown in Figure 5, the RHRS pressure remains below 660 psi for a period of time which is sufficiently long so that the operator may reasonably be expected to act. The operator is alerted by the over-.

pressure alarm on the RHRS, by a letdown line hi pressure alarm,,

and by a reactor coolant loop pressure indicator. The event is termi-nated once the operator turns off the pressurizer heaters. It is reasonable to assume timely operator action since the pressurizer heaters are in operation during that period of time during RHRS operation of high operator surveillance.

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FIGURE 1 RHR Loop - 10" RY 203 3ll 2

II To Pressurizer From Letdown ,Relief Tank Orifices 2" CYCS Li e 702 703 3/hI" To Letdown Containment

'enetration 4'112 HCY 133 II 2

CONFIGURATION OF RMRS'ELIEF LINE S-5 2/24/77

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FIGURE 2 Setpoint Pressure 200 Hater - Below 200'F

'ressure at 5A Blowdown (Maximum Blowdown) 180 '~~ Pressure

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10Ã at Accumulation (Rated Flow)

Hater 5 Steam - 380'F

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(Design Basis)

~t 120 RELIEF YALYE 60 PRESSURE vs. FLOH

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R. E. GINNA RV 203

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I 100 200 300 400 500 600 700 PRESSURE AT VALVE - PSIG

FIGURE 3 200 180 160 140 120 i cg I

100 80 FLOM vs..LINE PRESSURE DROP R. E. GINNA RHRS PRESSURE RELIEF LINE RHRS LOOP CONNECTION TO RV 203 40 20 0

0 10 20 30 40 50 .. 60 70 80 LINE PRESSURE DROP - PSID

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FIGURE 4 R>R SYS i4 0$'ERPP SC '9'>AT!,il rl RNR SYSiill OVER?Bi SS'(Z*T!CR,"~i TQ CECAYg P. AT RX PAGT 2 R'9 3 500. 00 4T5.00 w 450 00 cr 425. 00 4 CO. 00 375.00 350.00 325 00 300 00 D

D nD D D D C7 Ai ~ D TIME (SEC)

RRi? SYSTill GVERPRESS?:l?!2AT!CR 0< PCE RHR SYSTE)lIGYCPiPESSv'TtlAT!Ci CQC TQ QECAY HCA'l RCE PI,QT 4 RQR 3 SOO.GO SQQ 00

)QQ.OC 60Q.CO D

500. CG 6

400 GC D D D D D ~

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RHR SYSTEM OYr>>C SSUR!>ATJOQ TER RCE RCE~ P1% BO!L>4 - 30 CPN lET~"Val PLOi L RUS SOO. CU 475.00 450, CO 25 CO 4CO.GG 3!5.00 350 ",0 325 00 300 GO 4 O O O O O O O O

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CJ A O At ~ n TIME TSEC)

RRR SYSTEH OYERDRESSLR!2P>!2 i gn", P.;E RCE P2R f"Il!aR - 3 0 Crab 'TGQVaI PLOT 3 RU!i 1 e00.00 Cat GGO GC Ctl ata att CD 0 700.00 P

I an t~ eGO.CO CD O

CJ 5 500 00 Et CC 409 00 Cl O O oN O Ca CD C3 Cl CJ Pl C7 O ala an 1!ME {SEC)

RRR SVSTE:i OVER."RESSURflA lCY r ROE -. PL% b'}Llew". - 3 0 CP,". LETCGSW PLOT 5 RUN a 2CGO. 0 CJ Ti59 0 CD

'4P 1530 0 a I

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CC TOGO CO 750 GO O'C 5"3 CO 2'a va 250 0: 4 I

Cl O Ct O C%

8Ct ca Cl O

cat TlME (SEC)

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