ML17244A744
| ML17244A744 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/22/1979 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 7908090330 | |
| Download: ML17244A744 (64) | |
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ROCHESTER GAS AND ELECTRIC CORPORATION
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89 EAST AVZIsIUE, ROCHESTER, N.Y. 14649 LEON D. WHITE. JR.
VICE PRESIISENT TELEPHO IE AIII.'ACOOE TIC 546 2700 june 22, 1979 Mr. Boyce H. Grier, Director U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Region I 631 Park Avenue King of Prussia, Pennsylvania 19406
Subject:
IE Bulletins79-06A and 79-06A Revision I entitled "Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident" additional informa-tion requested during staff review of responses R. E. Ginna Nuclear Power Plant, Unit No.
1 Docket No. 50-244
Dear Mr. Grier:
Attached is the reply of the Rochester Gas and Electric Corporation to a request from members of the NRC Staff for additional information as a result of the Staff review of our letter dated April 28, 1979 which responded to IE Bulletins79-06A and 79-06A Revision 1.
Our responses have been prepared by a continuing Task Force at Ginna Station.
The questions we have received from the NRC Staff and our responses are numbered to correspond to the bulletins'ction items.
Very truly yours, L. D. White, jr.
Att.
xc:
NRC Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C.
20555
'F90809>> ~, ~ %Re>>3oo389
PROVIDE DATE FOR COMPLETICN OF YOUR REVIEW.
UPCN COMPLETICN OF YOUR EFFORTS WE REQUIRE 'IHAT YOU SUBMIT A
SUMMARY
OF 'IHE RESUL'IS INCLUDING REVISIONS MADE 10 OPERATING PKXZDURES.
IN ADDITION, SUMMARIZE IHE REVIEW RESULTS AND ACTIONS X%KEN WADIS REGARD 'lO 'HK NATURAL CIRCULATICN MODE OF OPERATION.
Substantial progress has been made in ccmpleting the procedure changes described in our April 28, 1979 letter.
In addition several new procedures have been developed.
RCS accident emergency procedures have been reviewed with particular attention being paid to the possibility of void formation.
Actions to prevent inadvertant formation of
- voids, as described in our April 28 letter, have been incorporated in the" procedures.
Specific emergency procedures which have been nullified include: E-l.l, Safety Injection Initiation; E-1.2, tuss of Coolant Accident; E-1.3, Steam Line Break Accident; E-1.4, Steam Generator Tube Rupture Accident.
A new emergency procedure, E-1.6, CVCS Breaks, is being developed.
'Ihis procedure will consolidate three existing procedures and will, we believe, provide for better operator response to breaks in the CVC system.
We expect this procedure to be ccxnpleted and approved for use by July 2,1979.
Two procedures dealing with heat removal frcm the primary system and potential void formation are being prepared.
Procedure 0-8, Natural Circulation, will provide instructions on identifying whether or not heat is being removed frcm the primary system by Natural Circulation.
Key parameters are identified and instructions for establishing, or reestab-lishing, natural circulation are provided.
Information fran the pre-operational tests which were performed at Ginna is employed as well as guidelines frcm Westinghouse.
We expect that this procedure will be ccmpleted ard approved for use by July 2, 1979.
In the interim, operator training sessions have
- covered, ard will continue to address, key features of natural circulation.
(see information fran lesson plan, Attachments 2, 3
a 4).
A new procedure, E-1.5 Void Formation in the RCS, has been written and is currently undergoing plant review prior to approval.
Final approval is expected by June 25, 1979.
This procedure will provide guidance on how to eliminate voids fran the RCS.
In addition, guidance on removing heat fran the RCS if neither forced nor natural circulation is available ard if the break flow in a LOCA is not sufficient to remove decay heat.
'inally, operator training sessions will continue to include presentations on the most recent information (see lesson plan outline Attachment 2) which is available frcm Westinghouse and elsewhere concerning void formation.
PK)VIDE SUFFICIEÃZ INFORMATICS AS K) WHICH LINES ARE OR ARE NOT ISOLATED SO THAT A CONCLUSION CAN BE REACHED AS 10 NHEQiER ALL LINES EXCEPT THOSE NEEDED FOR NEEDED SAFEZY FEATURE OR COOLING CAPABILITYARE ISOLATED.
INCLUDE A DISCUSSION OF THE OPERABILITY OF REACTOR COOLANZ PUMPS UNDER THIS ISOLATED CCM3ITION.
Table 1 lists all valves which are isolated (closed) on either a containment isolation signal or a containment ventilation isolation signal.
A containment isolation signal closes the valves shown as items 1 through 28 in Table 1, trips the containment sump pumps, ard initiates a containment ventilation isolation signal.
Contairxnent ventilation isolation closes the valves shown as items 29 through 38 in Table 1 and trips the purge supply 'and exhaust fans.
Remaining lines, Mich are not isolated on a containment isolation or containment ventilation isolation signal, are those required for safety functions.
In edition, ccmponent coolie water to the reactor coolant pumps and seal injection to the reactor coolant pumps are not isolated.
'Iherefore, the reactor coolant pumps remain operable under this isolated condition.
In our response of April 28, we stated that safeguard logic schemes were being reviewed in order to identify whether any changes were required with regard to containment isolation.
That review has been cmpleted with the conclusion that no changes were necessary.
We have further verified that all lines which penetrate containment that are not required for safety features or reactor coolant pumps are isolated by either locked valves,,
normally closed valves or autanatic valves that close on an isolation signal.
PROVIDE A SCHEDULE FOR COMPLETICN OF THE REVIEN OF OPERATING PROCEDURES MD TRAINING INSTRUCTIONS'NCORPORATING SUCH MODIFICATIONS AS ARE NECESSARY TO ENSURE 'IHAT OPERAK)RS WILL M7Z OVERRIDE AUNMATIC ACZIONS OF ENGINEERED SAFEZY FEATURES, UNLESS CONTINUED OPERATION OF ENGINEERED SAFETY FEATURES WILL RESULT IN UNSAFE PLANT CONDITIONS IN ORDER IO COMPLY WI'IH ITEM 7.A OF THE BULLETIN.
CLARIFY %HE MEANING OF WHAT IS MEANT BY REQUIRING CONCURK2lCE BY 'IWO LICENSED INDIVIDUALSTO OVERRIDE EMERGENCY SAFEZY FEATURES.
The following Emergency procedures have been reviewed, modified and issued as necessary to ensure proper operator actions.
a).
E-l.l. (Safety Injection System Actuation) b).
E-1.2 (Loss of Coolant Accident) c).
E-l.3 (Steam Line Break Accident) d).
E-l.4 (Steam Generator Tube Rupture)
Specific guidelines have been included Mich closely reflect the Westinghouse reccmnerdations.
The revised emergency procedures are currently being presented to the operators by the Training Department.
This presentation will be ccmpleted by July 13, 1979.
Also, changes to procedures are reviewed by licensed personnel via the procedure acknowledge book.
In addition, Mministrative Procedure, A-54.1 (Licensed Personnel Authority) was modified by Ginna Station procedure change notice No.79-1168 to state that two licensed operators shall agree on any overriding before the overriding action is executed on any safeguard system active component.
This administrative action is designed to meet the guidelines of IE Bulletin 79-06A in allowing the operator to override any canponent'n the safeguards system if the continued operation of that ccmponent will result in unsafe plant conditions while ensuring that undesired overrides, such as stoppage of SI flow to the, RCS when it is required for core cooling, is precluded.
YOUR RESPONSE K) ITEM 7.B APPEARS 10 BE INADEQUATE WI'Bi REGARD K) 'IHE REQUIREMEPZS OF ITEM 7.B OF THE BULLETIN.
PROVIDE ASSURANCE THAT OPERATING PROCEDURES WILL BE MODIFIED 1O KEEP HIGH PRESSURE INJECTION AND QiARGING PUMPS IN OPERATICN IN ACCORDANCE WITH THE CRITf~A SPECIFIED IN ITEM 7.B OF IHE BULU".TIN.
PROVIDE A SCHEDULE FOR CCMPLETICN OF THE REVIEN OF OPERATING PROCEDURE INCORPORATING SUCH K)DIFICATIONS AS ARE NECESSARY K) COMPLY WIIH ITEM 7.B OF THE BULLETIN The guidelines set forth by the NRC in the IE Bulletin 79-06A Rev.
1 for stopping SI canponents are not consistent with the guidelines that we have received fran our NSS supplier, Westinghouse.
It is RGGE position at present to closely reflect the Westinghouse recmmendations.
As stated in our response to item 2, procedures E-l.l through E-l.4 have already been revised.
In addition more information will be forthccming frcm Westinghouse, as a result of their response to a letter fran D.F. Ross, NRC to T.M. Arxlerson, Westinghouse, dated June 4, 1979.
Item 14 of that letter requests, in part:
Provide the results of an analysis of the effects of different HPI termination criteria on the course of small ZOCA's.
Specifically for each small break LOCA analyzed, ccmpare the effects of the NRC HPI termination criteria (as stated in IE Bulletins79-06A and 79-06A, Rev.
1 item 7 (b)), to those for the HPI termination criteria Mich have been recommended to licensees with Westinghouse designed operating plants.
Upon receipt of additional or revised guidelines fran Westinghouse which are applicable to our plant, with its lower head high head safety injection pumps, we will promptly revise the applicable procedures.
YOUR CRITERIA FOR TRIPPING REACTOR CQOLANZ PUMPS IS INCONSISTENT WITH 'IHE PROVISIONS OF ITEM 7.C OF %K BULLETIN.
PROVIDE ASSURANCE THAT OPERATING PROCEDURES WILL BE MODIFIED 10 KEEP REACTOR COOLANT PUMPS IN OPERATICN IN ACCORDANCE WIIH ITEM 7.C OF IE BULLETIN.
PROVIDE A SCHEDULE FOR GMPLETION OF 1'EVIEW CF OPERATING PBDCEDURES INCORPORATING SUCH MODIFICATIONS AS ARE NECESSARY Kl COMPLY WITH XTEÃ 7.C OF 'IHE BULLETIN.
The NRC requirements for RCP operation in an accident condition are not consistent with the Westinghouse reccaxnendation.
At present ROTE is using the Westinghouse reccmmendation for stopping RCP's in an accident situation.
Westinghouse's guideline is to stop all running RCP's &en system pressure is < 1550 psi and the SI system is delivering water to the RCS.
This recamnendation has been incorporated in the immediate actions of the Loss of Coolant, Safety Injection System Actuation and S/G Tube Rupture procedures (E-l.l, 1.2, and 1.4).
- However, 1500 psi has been used because the wide range pressure indicator on the main control board'as a major division at 1500 psig.
Therefore, this pressure is chosen as an aid to the operator.
In the Steam Line Break Accident procedure the RCP's are tripped immediately once the accident is identified to slow down the heat transfer rate to the steam generator.
'This condition is being discussed on an continuing basis with Westinghouse.
It should be noted that analysis of main steam line breaks show a rapid RCS pressure decrease with the pressure dropping below 1500 psia withing 25 sec for large breaks arxl within about 2 minutes for a break equivalent to a stuck open relief valve.
- Thus, we believe the criteria in this procedure are consistent with the Westinghouse guidelines.
PLEASE VERIFY THAT THE IDENTIFIED PARAMETERS HAVE BEEN INCLUDED IN APPROPRIATE OPERATING PROCEDUKX.
Pressurizer level has been eliminated in the diagnostic scheme for the identification of accident category.
Pressurizer level has also been removed as a symptan from all RCS accident procedures except CVCS break.
'Ihe operator is directed to other indications for accident identification.
The diagnostic tree is attached.
The indications listed in the initial K&E response to NRC bulletin 79-06A Rev. l were incorporated from the Westinghouse response to the bulletin.
Wide range RCS temperature a pressure Steam pressure Steam generator water level Containment pressure RWST leve1 Condensate storage tank level Pressurizer water level Boric acid storage tank level All the parameters listed are not indicative of RCS water inventory.
Each of these parameters except the condensate storage tank level are however, procedurally addressed in the IDCA and steam line break procedure.
The Boric acid storage tank level is addressed in DX'A and steam line break procedures in the SI pump suction change over verification to the RWST.
The remaining indications listed above are symptoms of the accident, are used during the accident as operator aids for verification of safe plant conditions, are required for operator action during the accident and/or are alarmed on the control board.
(The attached diagnostic tree provides an example of how the operator is instructed to consider a variety of plant indications in his evaluations)..
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l.iiw ov fnlfl>>g proaa<<rlKor prcaaiiro Al)>>ormn1 ly low nL'oain proaaiivo I)) o>>i! or 1)iitli nLcnni ge>>nvnL'i)ra I>>)l lento)) >>to>>in 1)renk Illal>>g or>>nrn)ni ntcnu) prcaa<<ro Ill I)otl) i)L'calli Qo>>orr)tora i)idIcal:ca loan of cooln>>t or tiibe r<<pt)lro Vct I.fy I)y cl)ockl>>II forl 1).
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S.l;.
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PLEASE PRQVIDE YOUR SCHEDULE FOR COMPLETING REVIEWS OF ALIGNMENT REQUIREMENTS AND PROCEDURES CONTROLLING MANIPULATION OF SAFETY-RELATED VALVES.
SUBMIT A
SUMMARY
OF THE RESUL'IS OF THE REVIEWS AND ANY REVISIONS NECESSARY WITHIN IWO WEEKS AFZER COMPLETION OF THE REVIEÃS.
ALSO REVIEW PLANT PROCEDURES AND REVISE 'IHEM AS NECESSARY 10 ENSURE THAT LOCKED SAFETY-RELATED VALVES ARE SUBJECTED 'IO PERIODIC SURVEILLANCE.
SUBMIT A
SUMMARY
OF 'IHE RESULTS OF IHE REVIEW.
All Periodic Test (PT) procedures that concern the Engineered Safety Features Systems have been reviewed ard updated where applicable, to ensure proper valve realignment following testing.
Each procedure directs test personnel to properly realign systems.
Further, additiional steps have been added to assure the systems have been realigned for operation.
This additional verification of realignment of systems is performed by personnel other than Testing Personnel, usually the Operation Department.
The Periodic Test (PT) procedures are performed on.a scheduled basis according to Technical Specification requirements, and prior to ard following Maintenance of Engineered Safety Features Systems.
The Periodic Test (PT)'rocdure changes were reviewed ard approved by the plant operating review ccmnittee (PORC).
Table 2 lists the PT procedures which have been revised to incorporate the additional steps and describes the changes.
New procedures have been developed and incorporated in Plant Operations to assure proper system valve line-up.
These system valve position verification procedures are performed on a regularly scheduled basis.
The performance of these procedures is in addition to the valve verification steps included in the Periodic Test procedures.
There are no Technical Specification requirements regarding locked valve surveillance.
An Administrative procedure, (A-52.2), governs the control of all locked valves at our facility, and states:
"The purpose of this procedure is to describe the requirements for a locked valve, authorities involved, documentation required and instructions relative to locked valve operation."
The purpose of locked valves at our facility is to provide control of equipment ard to maintain reactor safety, engineered safety features system alignment and personnel safety.
The shift foreman has the authority to issue a key to unlock a valve and the head control operator has the duty to maintain valve status, if changed, in the locked valve operations log.
Surveillance of these locked valves is accanplished by valve verification steps in Periodic Tests (PT's) following System Testing and by regularly scheduled system valve verification procedures noted below.
S-30.1 Safety Injection System valve position verification.
S-30.2 RHR S-30.3 Containment Spray S-30.4 Auxiliary Peedwater S-30.5 Standby Aux.
PW S-30.6 Safeguard valve position verification (inside CV)
PLEASE PROVIDE A COMPLETE RESPONSE Kl ITEN 9 AND IDENTIFY THOSE ISOLATION VALVES WHICH NAY BE REPOSITIONED AS A RESULT OF RESETTING CChVZAINMEÃZ ISOLATION.
Table 1 lists all lines which are designed to transfer potentially radioactive fluids fran contairxnent.
The containment isolation signal closes valves 1-28 (Table 1), trips the containment sump pumps, ard initiates a containment ventilation isolation signal.
Containment ventilation isolation closes valves 29 thru 38 (Table 1) and trips the purge supply ard-exhaust fans.
It should be noted that the resettirg of the SI signal does not reset contairxnent isolation or containment ventilation isolation ard therefore does not reposition any valves.
Valves (by number frcm Table 1) 3, 15, 16, 17, 19, 20, 21, 22, 23, 24, and 26 are the valves that are normally open and would return to the ogen position upon reset of Containment Isolation unless additional actions were taken to preclude this.
Note that number 19, 20, 21, ard 22 will not reogen if a high radiation cordition in the steam generator blowdown system exists.
Present LOCA ard Steam Line Break procedures require that the operators place all containment isolation valve switches in the "closed" position, prior to resetting contairxnent isolation.
Furthermore, emergency procedures have been modified to instruct the operator to place the contairment sump "A" pumps in the gull-stop position prior to resetting the containment isolation signal.
In the S/G tube rupture accident, it is our position that normal charging, letdown, spray ard power operated relief valve ogerations need to be established quickly ard that no radioactive transfer is possible frcm containment to the environs, except through the secondary steam lines.
Therefore, containment isolation is reset after the accident is identified as a S/G tube rupture ard the contairment vessel sump pumps are pull stopped, ard the letdown isolation ard reactor coolant pump seal return containment isolation valves are closed.
Prior to the
'HAMI Incident, there were several letters exchanged between RosE ard the NRC which dealt with safety actuation circuits ard their overrides.
'Ihese letters are provided in Attachment 1 to this letter for your convenience.
Furthermore, containment ventilation isolation does not autcmatically reset when SI or con-tainment isolation is reset.
'Ihe reset of containment ventilation isolation requires the use of a key which is held by the Shift Foreman.
Item 10 a.
PLEASE PROVIDE THE BASIS FOR THE CONCLUSION THAT EXISTING PROCEDURES FOR VERIFICATION OF OPERABILITY OF SAFETY-RELATED SYSTEMS ARE SUFFICIENT.
Fran the basis of the Technical Specifications Sections 3.3 and 4.5, the active canponents (pumps and valves) of safety related systems are to be tested aanthly to check the operation of the starting circuits ard to verify that the punps are in satisfactory running order.
'Ihe test interval of one month is based on the judgement that more frequent testing would not significantly increase the reliability (i.e., the probability that the ccmponent would operate when required) and would result in increased wear of long periods of time.
If a component is found to be inoperable, it will be possible in most cases to effect repairs ard restore the system to full operability within a relatively short time.
For a 'single canponent to be inoperable does not negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures.
To provide maximum assurance that the redundant ccmponent(s) will operate if required to do so, the redundant canponent(s) are to be tested prior to initiating repair of the inoperable component.
This is applicable to the safety injection pumps, residual heat removal pumps, and their associated valves (T.S.
3.3.1.2),
and to the containment spray pumps ard valves (T.S.
3.3.2.2).
%he two containment spray pumps also must be tested to demonstrate operability before initiating maintenance on an inoperable charcoal filter.unit (T.S. 3.3.2.2).
With one diesel generator out of service, the remaining diesel generator is run continuously provided such operation is not in excess of seven (7) days (total for both diesels) during any month, and provided the station transformer is in service (T.S. 3.7.2).
Testing of redundant canponents on the auxiliary feedwater system is not required by our Technical Specifications.
However, Maintenance ard Turbine Plant Operations procedures have been revised to"notify the Results and Test Department to test the redundant ccmponent(s) prior to initiating repair of the inoperable canponent.
Table 2 lists the procedures which have been revised to include this requirement.
Item 10 b.
PLEASE PMVIDE YOUR SCHEDULE FOR COMPLETING THE REVIEW AND MODIFICATION OF PROCEDURES.
WITHIN IWO WEEKS AFZER COMPLETING THIS EFFORP, PLEASE SUBMIT A SUM%6K OF THE RESULTS OF THE REVIEW AND 'lHE ACTIONS TAKEN.
Maintenance procedures have been reviewed and are sufficient to ensure the operability of safety related systems after they are returned to service following maintenance.
After maintenance, the Results and Test Department is notified and performs the applicable periodic tests to assure that the system is operable.
Periodic test procedures have been reviewed and changed as noted in Table 2.
After testing of a safety related system, an additional verification by non-test personnel is required.
All safety related valves manipulated during the test are checked to ensure they are in their proper position.
After this verification, the ccmpleted test procedure must be reviewed by the Head Control Operator ard approved by the Shift Foreman.
Item 10 c.
PLEASE IDENTIFY IHE LEVEL OF MJZHORITY REQUIRED EOR REMOVING AND RBZURNING SYSTEHS 10 SERVICE AND DESCRIBE THE METHOD USED FOR TRANSFERRING INFORMATION AMXJZ 'IHE STATUS OF SAFETY-RELATED SYSTEM AT SHIFT CHANGE.
The Shift Foreman must approve the removal from and the return to service of all safety related systems as required by Administrative procedure A-52.4, " Control of Limiting Conditions for Operating Equipnent".
A procedural
- change, approved on April 30, 1979, to A-52.4 also requires notification of the Head Control Operator.
The transfer of information about the status of safety related systems at shift change is accanplished thxough administrative procedure A-52.1, "Shift Organization Relief and Turnover", which lists the requirements for shift turnover.
Iten 12.
PLEASE PRDVEDE A SCHEDULE FOR WHEN PROCEDURES DEALIKG WITH HYDROGEN (RS Bl ZK PRIKQK CCOGVVZ SYSTEM WILL BE PREPARED.
A procedure to deal with hydrogen gas in the arimary coolant system is being prepared an9 will be approved by PORC for use by July 16, 1979.
4
2.
bXN-313 N3V-813
& 814 Seal water return isolation valve Supply & return ccmponent ccoling water to reactor support ccoling 3.
5.
6.
7.
.8.
9.
AOV-371 MV-539 AOV-846 AOV-951 A3V-953 AGV-955 ADV-966A Letdown isolation valve Pressurizer relief tank to Gas Analyzer Master N2 stop to Accumulator Pressurizer steam space sample Pressurizer Liquid space sample "B" Loop hot leg sample Pressurizer Steam space sample line isolation valve 10.
KV-966B R3V-966C
- Pressurizer Liquid space sample line isolation valve "B" Icop ?at leg sample line isolation valve 12.
13.
14.
15.
16.
17.
AOV-959 LCV-1003 A&B AOV-1600A mm-1721 M)V-1786 AOV-1787 RHR lcop sample valve 3A & 1B Reactor Coolant Drain Tank suction level control l
Reactor Ccolant Drain Tank to Gas Pzalyzer Suction line to Reactor Ccolant Drain Tank Reactor Ccolant "Drain Tank to Vent Header Reactor Coolant Drain Tank to Vent Header Secondary isolation valve 18.
19.
20.
AOV-1789 CV-70 CV-71 Reactor Ccolant Drain tank to Gas Analyzer "A" S/G blomicwn valves "B" S/G bloMown valves
21.
22.
23.
24.
26.
CV-76 CV-77 AOV-1723 PQV-1728 AOV-508 CV-74 "A" S/G blo~wn sample isolation valve "8" S/G bloMown sanple isolation valve Containment Smp Pump discharge stop valve Contairment Sump Pump Discharge'oot valve Reactor Makeup Hater to CV stop valve Instrument air ta, containment..isolation valve 27.
MV-8418 Demin. water to containment isolation 28.
H2 Reccmbiner solenoid valves.
VEÃZILATICNISOLATION VALVE CLOHJBE 29.
30.
31.
32.
33.
34.
35.
36.
37.
38.
%%&869 AOV-5870 AOV-5878 AOV-5879 ADV-1597 AOV-1598 AOV-7970 AOV-7971 i'-ATV1 Purge supply outside containment Purge supply inside contairment Purge exhaust inside contairment Purge exhaust outside contairnent Radiation nanitor supply valve Radiation monitor. e~ust valve Containment depressurization valve inside Containment depressurization valve outside Containment air test supply valve Tm contairznent air test vent valve
TABLE 2 PROCEDURE CHANGES PROCEDURE PT-2.1 Safety Injection System Pumps 1). Verification by non-test personnel that all safety related valves manipulated during the test have been returned to their required position.
2).
Head Control Operator in addition to Shift Foreman is notified before and after testing.
PORC APPROVED June 4
PT-2. 2 Residual Heat Removal System PT-2.7 Service Water System PT-2.8 Component Cooling Water Pump System PT-2.9.1 Check Valve Exercising Quarterly Requirement (RCOZ Pump Discharge)
PT-3 Containment Spray Pumps G NaOH Additive System.
PT-16 Auxiliary Feedwater System
TABEZ 2 PROCEDURE CHANGES PORC APPROVAL M-ll.5B Major Mechanical Inspection of APNP Notification of Results 6 Test Dept. to test redundant pumps before removing a pump fran service.
June ll M-ll.5C Minor Mechanical Inspection of*
A8%'-ll.5E Pipefitters Inspection of Motor Driven AM?
T-41B Turbine Driven AFWP Removal fran Service T-41D Mtor Driven AH@
Isolation M Maintenance Procedures PT-Periodic Test Procedures T Turbine Plant Operations Procedures
'X ROCHESTER G'$ ANO E'ECTRIC CORPORA 1ClV
~
c9 =".-'$I 'V HI., PC HEST""R,,V.Y. lA-'9' XD'I D WHIT'g. JR.
VICC PWCCICICIIT J<II<<M 2, 1979
,c~
> er' gO I
.Director of Nuclear Reactor Regulation Atiention:
Mr. Denn's L. Ziemann, Ch'ef Operating Reactors Branch No.
2 U.S. Nuclear Regulatory Commission Hashing on, DC 20555
Subject:
Containment Purging During Normal Plant Operat'cns R.
E.
l irma. Nuclear Power Plan Docket No. 50-244
Dear Mr. Ziemann:
This letter is 'n response to your let er dated November 29, 1978 which was received on December l, 3.978 rega ding elect 'cal bypasses and. overrides in the containment purge system.
Zt is our plan to justify limiting purging outl'ned in option 2 of your letter.
Our evaluation for justifying continua-tion of. limited purging during power ope at'on should be completed bv Ju3.y 2, 1979.
Pending complet'on of the NRC staff review o" that evaluat'n the R.
E. C'na,Nuclear Power Plant will limit purging to 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year while the "eac"or is crit'cal or operating as defined in the R.
E. pinna Technical Spec'catiors.
No time restrictions need be placed cn pu g'ng at shutdown cond'-
t'ons.
Zt should. be noted that limiting containmeni purging to 90 hr/yr result in inc eased pe sonnel ergosu~e during recuired Technical Specification surveillances.
A review of the conta~~.eni venti'at'on isolation sysi ~ has been made.
The conta'rwent venii'ai'on isola 'on system consists of the four containment purge valves, two containment depressuriza-tion valves and two rad'at'on mon'tor valves.
Zf open, these valves w'ill automatically close on a Sa ety Xnjec ion (SZ) s'gnal or on high containment ac ivity. If the containment vent'lation isolation system reset 's activated whi' a high containment activitv signal or Si signal is present these eight valves could be opened. and the automatic closure of these valves is blocked until the reset is deac 'vated.
The reset is deac
'vated when both the Si signal anc the high containmIent activity signals are cleared.
The purpose of the reset on the containmen ventilation isolat'cn system is to allow aurging o" containment 'n c=de to limit
a ROCHESTER CAS AHO ELECT~CORP.
January 2,
1%%
TO Mr. Dennis L. Ziemann SHEET HO.
potential hydrogen concentration buildup following a postulated LOCA when high containment activity and Sl signals could be present.
Procedures associated with the activation of 'the containment ventilation isolation system reset have been modified to alert the operator that activating the reset blocks automatic closure of'he eight valves on an SI signal.
Zf a high containment activity alarm is present the reset should not be used until the high containment'ctivity alarm has been cleared unless SZ has occurred.
A review of all'emaining safety actu'ation signal circuits which incorporate a manual override feature is in progress.
This review should be completed by mid February 1979.
Until this review is complete the use oz bypasses on unreviewed circuitry will be minimized to the maximum extent possible.
Zt may, however, be necessary in certain instances to employ overrides'r resets in order to perform certain necessary operations such as instrument tests or equipment maintenance.
Very tzuly yours, L. D.
$"hi
, Jr.
ROCHESTER GAS AND ELECTRIC CORPORATION
~ 89 EAST AVENUE, ROCH STER, N.Y. I ~6~9 C f ~ SIC a ~ ta c cga 5'62'00 February 16, 197PQ+gQ g Director of Nuclear Reactor Regulation Attention:
Mr. Dennis L. Ziemann, Chief Operating Reactor Bran'ch No.
2 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Review of Safety Actuation Circuits with Overrides R.E. Ginna Nuclear Power Plant Docket No.
50-244'ear.
Mr. Ziemann:
Your letter dated November 29, 1978 requested restrictions be placed on containment purging during normal operation and that a review of all safety actuation signal circuits which incorporate manual override features be made.
The Rochester Gas and Electric Corporation (RG&E) letter dated January 2,
1979 transmitted the RG&Z commitments on purging during normal operation and stated that the required review of override features would be completed by mid February 1979.
The purpose o
this letter is to transmit the results of that review.
Details of the rev'w are presented in Appendix A.
The review of a'll safety actuation signal circuits which incorporate a manual override feature indicates tha" actuating a
particular override.(reset) does not cause the bypass of other
, safety actuation s-'gnals.
The reset switches described in Appendix A are push button switches located on the control board wi&.".o physical restraints..
I In all cases where the safety actuation signal is generated automatically and the reset switch is actuated, the safety actu-ation signal will be 'nhibited until all logic paths for auto-matically generating the safety actuation signal nave opened.
Once.all logic paths
- open, the par icular safety actuation signal reset relays de-energize and e-establish the ability to auto-matically genera" e th s.
ety actuatio~. signal.
S'nce the rese+
remains actuated only if the input signals causing the automatic safety actuation signal pers'st and these input signals are annunciated, no s'eparate annunciation for the reset actuation is necessary.
The operator has sufficient information "o deduce a
certain system is in the reset moce.
ROCHESTER GAS AHO ELECTR~!CORP.
O~TE February 16, ~79 To Mr. D.L. 2ienann, Chief 1
SHEET NO.
ln no case does actuation 'of a particular reset s~sitch prevent the operator from manually operating the equi"ment from the control board.
Therefore, operation of a reset does not preven equipment fror operating which is necessary to mitigate the consequences of a postulated acc'dent.
V'ry truly yours,
~~+.
L. D. Nhite, Jr.
LDH:np Attachment
Appenaix A
,Review of Safety Actuaticn Signal Circuits Incorporating Manual Overrides The following summarizes the results of a review of safety actu-ation signal circuits which incorporate a manual overriae feature.
The purpose of the review is to ensure that overriaing o
one safety actuation signal does not also cause the bypass of any other safety actuation signals:
1.
SAFETY INJECTION CIRCUIT:
This circuit has a reset switch which gives, the operator the means of resetting safety injection one minute or longer after initiation.
Actuation of the reset swatch in itself does not change the state of any eauigment, but permits the operator to place the ec;uipment affectea by safety in]ection to the position desirea.
ll If safety in)ection is caused by'utomatic actuation, and the reset switch is actuated, automatic safety ingection will be inhibited until all.logic paths for automatic safety inqection have openea.
Once all logic paths
- open, the safety injection reset relays de-energize ana re-establishes
.automatic safety injection cal abilities.'anual safety injection initiation is available at all times.
There is no arnurciation of the safety zngection circuit being in the reset moae.'he purpose of the reset switch on the safety inqection system is to allow ecuipment t'o be realignea for the recir"ulation phase of a postulated LOCA.
2.
CONTAINMENT VENTILATION ISOLATION CIRCUIT:
This circuit has a reset switch whicn gives tne ol;erator
~ the means of resett'ng containment vertilation isolation.
Once the reset switch has been actuated, most of tne ecuipment will automatically return to the state selectea prior to the isolatior. signal.
If containment ventilation isolation was caused automatically, either by safety in)ection or high raaiation alarm on containment gas and/or particulate monitors, ana this cordition continues to ex'st a te the reset switch has been actuated, ther. containment ventilat'on iso'ation cannot be achievea automatically or by the manual isolation switches until th' logic clears.
Once the automatic logic clears, the containment ventilation isolation reset relays de-en< rgize ana re-establishes automatic or manual isolation capabilities.
.Manual operation of the valves from the control board 's available at all times.
There is no annunciation of the automatic containment ventilation isolation system being in the reset mode.
The purpose of the reset switch on the containment venti-lation isolation system is to allow purging of containment in order to limit potential hydrogen concentration buildup following a postulated LOCA when high containment activity and safety injection signals could be present.
3.
CONTAINMENT ISOLATION CIRCUIT:
'This circuit has a reset switch which gives the operator the means of resetting containment isolation.
Once the reset switch has been actuated, some equipment will return automatically to the position selected prior to the iso-lation signal.
If containment isolation was caused automatically by an automatic safety injection signal, and containment, iso-lation reset switch is actuated without resetting safety injection, containment isolation cannot be obtained by the manual containment isolation switches until safety injec-tion is reset.
Actuation of the -reset permits the operator to place the valves affected by the containment isolation signal in the position desired.
This capability is necessary so that the operator has flexibilityin dealing with post accident conditions within containment.
There is no annunciation 'of the automatic containment isolation being in the reset mode.
4.
CONTAINMENT S PRAY CZRCUZT:
This circuit has a reset switch which gives the operator the means o
resetting containment sp ay.
Once the reset switch has been actuated the spray additive tank discharge valves will return automatically to the position called for by the controller prior to the containment sp ay signal.
The containment spray pumps and their discharge valves would require operator action to change state.
If conta'nment spray was caused automatically by the high containment pr essure logic, and this logic continues to exist after reset, containment spray cannot be init'ated by the manual spray switches.
Once the high p essure logic has cleared, the containment spray eset relays de-energize and re-establishes au omatic or manual contain-ment spray capabilities.
Actuation of the reset permits the-operator to place the valves and pumps affected by the containment spray signal in the state desired.
This capability is necessary so that the operator has flexibilityin dealing with post accident conditions within containment.
The e is no annunciation of the automatic containment spray system being in the reset mode..
5.
FEEDNATER ISOLATION RESET:
This circuit has a reset switch which gives the operator the means of resetting the isolation signal. to the 'feed-water bypass valves only.
The main feedwater valves will remain closed until the isolation logic clears, and then they automatically assume the position reauested by their control circuit.
If feedwater isolation is caused by high steam generator level logic, and this condition still exists after the reset switch is actuated, a safety injection signal would not cause an isolation to that particular feedwater bypass valve. It should be noted that a safety injection signal also causes the main feedwater pumps to be tripped, therefore, closing the feedwater bypass valves on a safety injection signal is redundant.
There is no annunciation of the automatic feedwater iso-lation system being in the reset mode.
6.
NUCLEAR INSTRUM NTATION SYSTEM DEFEAT,
- BYPASS, AND BLOCK SNITCHES:
This system has several switches which are used for the following purposes:
(a)
Defeat Switches - Defeats a permissive which rein-states a
rip logic.
(b)
(c)
.Bypass Switches' Bypasses a trip or runback function for calibration or maintenance purposes.
Protection is still provided by redundant channel or channels.
Block Switches Blocks trips generated by source, intermediate, and power range channels.
These switches are actuated as permissive setpoints are reached to permit taking reactor critical and up in power.
These blocks automatically reset as power is decreased below its particular setpoint.
All the above switches i" actuated, are indicated by one or more of the cllowing: status light, alarm on the computer, or actuate an annunciator.
7.
IHSTRUHENT &1D CONTROL DEFEAT SWITCHES:
The following switches and their circuits were reviewed to insure that they are only performing their intended function, and no other safety functions are being bye assed.
The purpose fo" these switches 's to be able to switch control from one sensor loop to another for testing, calibration and maintenance purposes.
In all cases, reactor trip and safety injection signals are generated prior to defeat
- switches, and are not affected bp switch position.
(a)
P/429A Pressurizer Pressure Selector Switcn Used to select two of the four pressurizer pressure channels for controlling pressurizer
- heaters, sprays, and power relief valve PCV-430.
(b)
L/428A Pressurizer Level Selecto'r Switch - Used to select two of the three pressurizer level channels for controlling charging pump speed, letdown iso-
- lation, and pressurizer heaters.
(c)
T/405K and T/405F Delta T Defeat Switches Used to defeat a channel from the over temperature and over power turbine runback circuit, and to remove.
a channel Delta T signal from the input of the summer for generating the average Delta T signal or the Roc Insertion Limit Circuit.
(d)
T/401A and T/401B Tavg Defeat Switches - Used to defeat a Tavg channel from the input to the average Tavg summer which is used for full length rod control, condenser steam
- dump, and pressurizer level setpoint.
t PyipiZl7
,I A%0 I//j~q
- l'ii')AS"I hK'Mid'OCHESTER GAS AND EI ECTRIC CORPORATION
/
~EO
~NI %\\lath I~
~
89 EAST AVENUE, ROCHESTER, H.Y. 14649 LEON O. WHITE. JR.
VICC tSCSIOtm' E LC I+~0 H C i~cd <<ooc -.ia 546 2700 March 30, 1979
~~e<
Director of Nuclear Reactor Regulation Attention:
Mr. Dennis L. Ziemann, Chief Operating Reactors Branch No.
2 U. S. Nuclear Regulatory Commission Nashington, D.C.
20555
Subject:
Review of Safety Actuation Circuits With Overrides R.E.
Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Ziemann:
Your letter of November 29, 1978 requested that we perform a
review of all safety actuation signal circuits which incorporate manual override features.
Our letter of February 16, 1979 pro-vided the results of that review.
The purpose of this letter 's to supplement and amplify the information on the containment ventilation isolation circuitry which has previously been pre-sented.
The containment: ventilation isolation circuitry and the reset function were described in our previous letter.
The isola-tion signal is generated either by safety injection (SX) or high containment radiation alarm on containment radiation monitors.
Once a signal is generated, the isolation function is locked in and can only be cleared through use of the reset function, even if the initiating signal has been cleared.
- Further, once a high radiation signal is generated, the signal itself is locked in and must be cleared.
This is not clea ed by the containment ventila-tion isolation reset.
As stated in our previous letter the reset function is not annunciated.
However, the signals that generate a containment ventilation isolation signal are annunciated.
The position of the purge valves is indicated by lights on the control board.
- Hence, a combination of an isolation signal (annunciated) and a
lack of a close light for the respective valve is positive indica-tion the reset function has been actuated.
The reset switch is a key lock switch and i"s use is covered by strict administrative controls.
Situations in which it might be required are as follows.
Zf a spurious S.Z. or high radiation signal were to generate an. isolation signal, it would be necessary
ROCHESTER GAS AND ELECT~CORP.
o~TE March g0, 1gW>>
Mr. D.L. niemann, Chief SHEET NO.
to use the switch to clear the isolation signal.
Before using the reset,
- however, the plant operator would clear the spurious signal.
- Thus, when the reset was employed, it would momentarily block the signals but following release would not block subsequent signals.
Procedural precautions alert the operator to the fact that the spurious signal should be cleared prior to using the reset.
Strict contxol of the key for the reset under the Shift Foreman ensures that proper procedures are followed inasmuch as no single operator error can result in improper use of the reset function.
A second situation involving the use of the reset key switch is following the monthly test of the containment ventilation isolation circuitry.
Zn this test, a simulated signal is input into the circuitry.
Following completion of the test, the test signal is removed and cleared.
Only after this is accomplished is the isolation signal cleared, again under strict administrative controls including decisions by two operators.
A third circumstance which could involve use of the key switch is an actual high radiation signal which isolates contain-ment when purging is desired.
Purging could be accomplished by use of the reset function thereby overriding the high radiation
- signal, however., this is not permitted without a detailed evalua-tion.
Zn addition, to the best of our knowledge, this has never occurred in nearly ten years of plant operation.
The practice, enforced by procedure, in this case is to attempt to clear the high radiation signal in case it is a spurious signal.
Zf it is not a spurious signal, the set point of the monitor would be evaluated and raised, while ensuring that all regulatory require-ments for release concentrations (e.g.,
3.0 CFR Part 20 limits) are met.
This would permit the high radiation signal to be cleared.
Once the high radiation signal were cleared, the ventila-tion isolation signal could be cleared by momentary use of the reset key switch.
Plant procedures for this situation, will explicitly provide information as to the function of the xeset function and the need to thoroughly understand and evaluate the situation at hand before using the reset.
Again, it has never been necessary, to the best of our knowledge, to use the reset to override an isolat"on signal in oux nearly ten years of plant operation.
Finally, it may be necessary to use the reset function in order to purge containment to limit hydrogen buildup in containment following a design basis loss of coolant accident (LOCA).
Follow-ing a LOCA, both a high radiation and SZ signal will exist.
Zf, based on. hydrogen sampling of the containment atmosphere, it is necessary to purge, the plant operator is provided detailed precautions on use of the zeset.
He is directed to place all valve position controllers in the close position so that no valve will open on initiation of the reset.
Then the operator actuates m
I
, ROCHESTER C4$ 4HD ELECTRI~ORP.
D4TE March 30, 197M 7;o Mr. D.L. Ziemann, Chief II I
SHEET NO.
the reset.
Finally, he initiates containment purge. It should be noted that purging to control the post-accident hydrogen concentration is not necessary until at least several days after the evenf.
In conclusion the Ginna containment ventilation isolation circuitry and procedures regarding its use are adequate.
This is based on the detailed procedural controls which have been imple-
- mented, the physical control of the reset key switch which involves at least two operators to use, and ten years of successful plant operation.
Very truly yours, Leon D. %hite, LDfv:np
ASSUi~GTlCHS:
ii1Il3zlUm~ eguards ecuL~ent s~Mg ~ of of s1te powerg
.4 lcop ~t, cold le. br aks.
BREAK SIZE
<Z/8" (No SI)
S/G throL@h forced or natural recxLco H3ZCQK PRESSK3E
)3/8 tt~lII (SI-1715) '/G throL~h natural cir-culation a=ter 1 day-core boiloz Scme in core
& ~
~alizes above S/G relief valve setting NOZZ:
Systm w~l fillsolid
)1 ffg2ff (SI-1715)
S/G thmLgh condensation when SI = break flow core boilof Throughout Tc)Tsat no natural chic~ due to voids 2" break 1/2 core for~2 min. KT-1000'F.
@wctu'3 l ~les belcw S/G relief set~
NOZE=
S/G can because a heat sourc I
)2ff S/G through conde~tion (ST initially thea core boil-1715) oz" when SI = break NCGZ:
S/G becomes a he t sourc
. Accumlulator diam NroLglxlut 4" break over 1/2 core for 8 min.
6" 80%
for 2 min.
PCT-1750.
Depends cn break size PGRV No heat sink Core boilof= can't ~I. ve the decay heat Nr the f2ZSt. l2lf lour ~
Throughout A=ter 75 min. 3'or 7 miIlo( 1 for 15 min.
Rma>m at 1500 for 1.S hours NCZE:
PZR. rVi. pliable PORV Aux P~
in 30 S/G ard Core boiloff NCPr:
PZR. r>L.
unrel~le'zougbout Stabaliz s
less than nda~
relief SO min.
,0 Cl t'
gpss
'+tel
~ >re r.a
+T Ni NY v
0 0
u C
4l el4 l
ill a 0+V 0
cv 0 X
Y OY nl C
'4
INITIALNATURAL CIRCULTION TEST 1.7$
3.4A 5.1%
6.8%
8.5%
TKVG.
550 562 25.5 40 4% actual 5.25% actual 5.7% projected 6.5% projected 6.9% projected
CC~W~
4