ML19224B840
| ML19224B840 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear, Crane  |
| Issue date: | 05/04/1979 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML19224B836 | List: |
| References | |
| 1-059-8, 1-59-8, NUDOCS 7906270041 | |
| Download: ML19224B840 (1) | |
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NSIC mani AAKANSAS PCWER S UGHT COMPANY N W 'Cs s c x =:i1 UTTLE ROC 4 AFma 72203 C5C1) 371-40cc May 4,1979 l-059 8 Mr. K. V. Seyfrit. Director i
Offfee of Inspection & Enforcement U. 5. fL c' e r Regulatcry Ccamission Region IV 611 Ryan Pla:a Drive, suita 1000 Arlington, Texas 76011
Subject:
Arkansas Nuclear Ona - Unit 1 Docket No. 50-313 License ?&
DPR-51 IE Bulletin.')-053 (File:
1510.1)
Gentlemen.
Attached is our reponse to IE Bulletin 79-053.
Very truly yours, O Q 4 M (. '/ d b David C. Trtable Manager, Licensing DCT:ew Attach:nent ec:
Mr. John G. Davis. Director (w/a)
Nuclear Regulatory Cc=nission Office of Inspection and Enforcenent Divisien of Reactor Operations Inspection Washington, D.C.
20555 Mr. W. D. Johnson U. S. Nuclear Regulatory Cenmission P. O. Ex 2090 Russel1ville, AR 72501 "t54 10 790627004/
Question:
1)
Develop procedures and Lrdin operac1on personnel on methods of establishing and maintaining natural circulation.
The pro-cedures and training must include means of monitoring heat removal officiency by availetsic planL Ins trumenta c1on.
Ine pro-cadures must also contain a method of assuring that the prirary coolant system is subcooled by at least 500F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory filling of the OTSG prior to securing the reactor coolant pumps, a detailed analysis should be done to provide guidance as to ths expected system response.
The instructions should include the a,
following precc9tions:
maintain pressurizer level sufficient to prevent loss of a.
level indication in the pressurizer; b.
assure availability of adequate capacity of pressuri:er heaters,, for pressure control and maintain primary system pressure to satisfy the subccoling criterien for natural circulation; maintain pressure - temperature envelope within Appendix G c.
limits for vessel integrity.
crucecures ano craining shall also be provided to maintain core cooling in the event both main feedwater and auxiliary feedwater are lost while in the natural circulation ccre ecoling mcde.
Rescanse:
Procedures are currently being revised tc provide greater detail in methods of establishing and maintaining natural circulation involving cases of unplanned total loss of fcreed circulation, anticipated loss of forced circulation and piant cocidewn utili-zing natural circulation.
These procedures shall include the follcwing guidance in assessing heat removal efficiency and natural circulation stability:
1.
Verification that foliowing loss of forced circulaticn frem rated pcwer, Reactor Coolant System (RCS) hot leg temperatures stabilize after reactor coolant pump (RCP) coastdcun and then tE nd to deC7 Gase, ani Ut RC3 bot leg' ' tema tends to decrease (erature and Cold leg temperature diffarr ^
i.e., tend to c::n-verge).
2.
Verification
-to-sacendary heat rejection is oc:ur-ring. by cbsem..s %t cud:ine bypass or atmospheric dmo valve cperation is required er steam generstar (OT5G) steam safety valve cparation is required to limit secondary pressure, and emergency feedwater (EFW) is required to maintain 075G level.
254 10
In the event conditions occur which force the operator to purpose-fully terminate all RCP operation (other than during a normal plant cooldcwn), procedures will be revised to direct that prior to stopping all RCP's:
1.
Flew to each OTSG is verified by placing EFW in operation, cycling the feed valves to each OTSG, and noting the cor-responding EFW pump discharge pressure change.
2.
Pressurizer level is verified to be between 50 inches and 90 inches.
3 RCS is verified to be at least 50 degrees subccoled.
N 4.
At least one bank of pressurizer heaters are demonstrated opera tional.
5, OTSG operate range level is to be slowly increased, commen-si, rate with the ability to maintain RCS pressure and pressur-izer level, to approximately 50% indication via normal feed-water (if normal feedwater is available and times permits).
Procedures currently require maintenance of RCS pressure-temperature relationship within Appendix G limits.
Proposed revisicns shall also contain such requirements.
Caution notes will be included in natural circulation procedures to require manual High Pressure Injection (HPI) initiaticn and operation of at least one RCP per loop (if cperable) in the event heat rejection via the OTSG's is i-rreparabl actions do not create an unsafe condition) y lost. (provided these Such procedure modifications and associated training on those modified procedures shall be complete prior to exceeding 5% rated pcwer frcm the current refueling outage.
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Ques tion:
- 2) Mudify the actions required in Item 4a and 4b of IE Bulletin 79-0EA
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to take into account vessel integrity considerations.
"4.
Review the action directed by the operating procedures and training instructions to ensure that:
a.
Spara tnet do not override automatic actions of engineered safety ' features, unless continued oceration of engineered safety features will result in unsafe olant concitfons.
For example, if continued coeration of encineerec safety features would tnreaten reactor vessel integrity then tne HPi shculd be secured as notec in o(2 celow).
-s b.
Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure conditien, it must remain in operation until either:
(1)
Both low pressure injection (LPI) pumps are in operatien and ficwing 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 in operation.fer 20 minutes, and all hot and cold leg temperatures are at least 50 degrees belcw the saturation temperature for the existir.g RCS pressure.
If 50 degrees subccoling cannot be maintained after HpI cutoff, the HPI shall be reactivated.
The degree of subcoolino bevond 50 decrees F and tne lenath of time Hpr is in oceration shall be limited by the cressure/
temcerature cons 1cerations for tne vessel intecrity"
_Resconse :
Our responses to question 4a and b in our letter of April 11,1979, are revised as folicws.
4a.
A caution note will be added to the ANO-1 emergency pro-cedures instructing the cperators not to override the aut -
matic actions of the Engineered Safeguards (ES), unless continued operation will result in an unsafe plant con-dition, without first determining the consequences of that override and consulting with the shift supervisor.
The precedures will also be mcdified to add clarifying steps to aid operatcrs in reccgnizing a spuricus actuation and provide for orderly termination of the supricus actuatien.
25A UI 4b.
The ANO-1 emergency procedures will be modified to specify the following actions, system has been automatically actuated because of a lowIf pressure condition it must remain in operation until:
1)
Law Pressure Injection is in progress with a ficw rate in excess of 2000 gpm and the situation has been stable for 20 minutes; or, 2)
The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50F belew the saturation temperature for the existing Reactor Coolant.:
System pressure.
If 50F degrees subccoling cannot be main ~
tained after HPI cutoff, the HPI shall be reactivated.
The degree of subcooling beyond SOF and the length of time the HPI is in operation shall be limited by the pressure /
temperature considerations for the vessel integrity; or, 3)
The RCS pressure returns to normal operating pressure with the temperature in the hot and cold legs being controlled with at least 50F subcooling by an operable
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Question:
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 PORY during anticipated transients.
This analysis shall include con-sideration of a modification of the high pressure scram setpoint and the PORV opening setpoint such that reactor scram will pre-clude opening of the PORV for the spectrum of anticipated tran-sients discussed by B&W in Enclosare 1.
Changes developed by this analysis shall not result in increased frequency of pres-surizer safety valve operation for these anticipated transients.
Resoonse:
- i.
We have discussed the opening of the Electrematic Relief Valve
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(ERV) during normal transients with B&W and believe the changes pr.oposed in our April 24, 1979 letter are appropriate to preclude lifting of the ERV during normal tr;:nsients.
Duririg our current refueling outage and befora exceeding 200F Reactor Coolant System (RCS) bulk system temperature (Tave) we will increase the relief setpoint of the ERV from 2255 psig. to 2450 psig and decrease the high RCS reactor trip setpoint from 2355 psig to 2300 psig.
These changes will preclude opening of the ERV during normal plant operating transients yet will not increase the expected frequency of safety valve operation.
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Ouestion:
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 cociant system.
These transients include:
a.
loss of main feedwater
, b.
turbine trip:
Main Steam Isolation Yalve closure c.
d.
Loss of offsite pcwer
- s e.
Low OTSG level f.
low pressurizer level Restonse:
Prior to exceeding 200F Reactor Coolant System (RCS) bulk system temperature (Tave), we will provide procedures and operator training for a prompt manual reactor trip, as appropriate, for the transients listed below.
sa. Loss of Main Feedwater - As committed in our May 3,1979 letter to Mr. Denten, we will provide a con ~ trol grade antici-patory reactor trip upon loss of main feedwater.
Therefore, precedures will require verification of reactor trip upon loss of main feedwater.
The tenn verification as it will be i
used in this procedure requires the operator to take the necessary manual actions to trip the reacter if the auto-matic system has not already caused a ' trip.
4b. Turbine Trip - As committed in our May 3,1979 letter to Mr. Denton, we will provide a central grada anticipatory reactor trip upon a turbine trip.
Therefore, procedures
-will require vefificaticn of a reactor trip.
4c. Main Steam Isolation Valve Closure - A manual reactor trip will be initiated by the operator in the event or a main
/ { [l u steam isolation valve closure.
4d. Loss of Offsite Pcwer - Upon separation of the Arkansas Nuclear One - Unit.1 (ANO-1) turbine generatcr from the offsite grid, the plant is desiened to runback to apprcxi-mately 15% Full Power and supply internal house leads via the main generater through the auxiliary transformer.
The auxiliary transformer is. fed directly fran the main generat:r and is the normal scurce of onsita pcwer fcr hcuse leads during operations.
Should the turbi. e trip during runback of g
the plant, the reacter wculd trip.
gSee cur response to Itra 46. above).
1 O
We believe it would be more prudent to allcw the plant to runback to approximately 15% power and supply hcuse loads through the auxiliary transformer than to trip the reactor upon grid separation and thereby lose an available source of power.
By maintainir.g the turbine generator on line, we are able to run all Reactor Coolant Pumps, as well as all safety and secondary equipment withnlit challeming the Diesel Generators and maintaining them as backup sources of power.
We therefore, state that we believe tripping t a. reactor upon separation of the turbine generatcr frca the system grid is a reduction in the margin of safety and such a trip
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shculd not be incorporated on ANO-1 4e. Low OTSG Level - OTSG startup range level indication varies in an approximately linear fashion with pcwer level.
As such, the steam generator icw level setpoints would require a sliding function with power level.
However, since loss of feedwater transients typically occur below 40% pcwer, a low level steam senarator trip of 15 inches will provide opt 1n n protection in the pcwer range of highest protability.
Above 40% pcwer, the reactor would be tripped on loss of feedwater indications, although the 15-inch administrative limit will remain active as a backup trip.
Therefore, above 40 power,.the reactor will be tripped upon OTSG level falling to 15 inches (indicated) or below.
4f. Low Pressurizer Level - Reactor trip upon low pressurizer level'is inconsistent with this question's general concern for overpressurization events, as a low pressurizer level is associated with underpressurization rather than over-pressurization.
- Further, a reactor trip in an underpres-surization (overecoling) event only adds to the problem as it removes the primary heat source. Current procedures require increasing reactor coolant makeup flew on decreasing pressurizer level, isolating letdown flow, and increasing seal injection ficw in order to return pressurizer level to normal.
If necessary, additional makeup flow ctn be obtained by opening the-high pressure injectica motor operated valves or starting an additional makeup ptrp.
If pressurizer level continues to decrease, exhibiting loss of coolant accident indications, precedures require reactor trip.
The reactor core will continue to be protected by the low RC pressure trip as analyzed in the FSAR.
254 175 Question:
5)
Provide for NRC approval a desion review and schedule for impie-mentation of a safety orade automatic anticicatory reactor scram for loss of feedwater, turbine trip, or significant recuction in steam generator level.
Resconse:
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- As commented in our letter cf May 3,1979, to Mr. Denton, during our current refueling outage and before exceeding 200F Reactor Coolant System (RCS) bulk system te.,perature (Tave), we will implement a control grade anticipatoiy trip to trip the reactor on a loss of Main Feedwater (MFW) or on a Turbine Trip.
Com-mensurate with the above time frame, we will submit a schedule for upgrading these two reactor trips to safety-grade trips.
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non e t.taa.
- 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 1n not in a controlled or excected cond_1 tion of oceration.
- Furthir, at tnat time an open continuous communication channel shall be estac11sned and maintained with NRC.
Resconse:
Our response to question 12 in our letter of April 16,1979, is hereby modified to include the following paragraph.
"In addition, upon notifying the NRC that such a condition exists, a telephone nunber will be provided to the NRC that can be reached on a centinuous (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day) basis until normal conditions are restored."
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question:
7)
Procese chances as recuired, to those ' technical specification's encn must be modifiec as a result of your imolementing tne atfove items.
- Rescense:
Our letter to Mr. R. W. Reid of April 24, 1979, proposed a change to the Arkansas Nuclear One-tJnit 1 (ANO-1) Technical Specifications to incorporate our response to Item 3.
We are currently evaluating the necessity of other changes to the ANO-1 Technical Specification's that will be required by implementing the above responses.
If further changes are required, we will p. wide the appropriate changes by May 21, 1979 254 1/8