ML19256F172
| ML19256F172 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/07/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| References | |
| NUDOCS 7911210571 | |
| Download: ML19256F172 (13) | |
Text
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+* p nasceq#o UNITED STATES 8
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NUCLEAR REGULATORY COMMISSION y
WASHINGTON, D. C. 20535 November 7, 1979 Docket Nos.: 50-317, 318 Mr. A. E. Lundvall, Jr.
Vice President - Supply Baltimore Gas & Electric Company P. O. Box 1475 Baltimore, Maryland 21203
Dear Mr. Lundvall:
We have reviewed the information provided by your letters dated April 26, May 8, June 5, and August 20, 1979 in response to IE Bulletin 79-06B for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
The enclosure provides our evaluation of your responses with respect to their specificity, complete-ness, and responsiveness to the intent of said bulletin.
In this regard, we have found that you have taken the appropriate actions to meet the requirements of IE Bulletin 79-06B.
It should be noted that the staff review of the Three Mile Island, Unit No. 2 accident is continuing and other corrective actions may be required at a later date. For example, IE Bolletin 79-06C was issued on July 26, 1979 requiring new considerations for operation of the reactor coolant pumps following an accident. Our review of Combustion Engineering's response to Items 2 and 3 of Bulletin 79-06C (Report CEN-ll6-P) and your responses dated August 20 and 30, 1979, is continuing pending the submittal of the long-term actions required by the bulletin.
In addition, new requirements may result from our generic review of procedures for operating C-E plants, our review of plant performance during feedwater incidents and small-break LOCAs, and from licensee's responses to the requirements delineated in NUREG-0578 and.NUREG-0585.
Sincerely, b
Q Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors
Enclosure:
Evaluation of Licensee's Responses to IE Bulletin 79-06B
!b cc w/ enclosure: See next page 7011210 i
Baltimore Gas & Electric Company cc:
James A. Biddison, Jr.
Mr. R. M. Douglass, Manager General Counsel Quality Assurance Department G and E Building Room 923 Gas & Electric Building Charles Center P. O. Box 1475 Baltimore, Maryland 21203 Baltimore, Maryland 21203 George F. Trowbridge, Esquire Shaw, Pittnan, Potts and Trowbridge 1800 M Street, N.ll.
Washington, D. C.
20036 Mr. R. C. L. Olson Baltimore Gas and Electric Company Room 922 - G and E Building Post Office Box 1475 Baltimore, Maryland 21203 Mr. Leon B. Russell, Chief Engineer Calvert Cliffs Muclear Power Plant Baltinore Gas and Electric Company.
Lusoy, Maryland 20657 I
Bechtel Power Corporation ATTH: Mr. J. C. Judd Chief Nuclear Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Cembustion Engineering, Inc.
ATTH: Mr. P. W. Kruse,ibnager Engineering Services Po:t Office Box 500 Windsor, Connecticut 06095 Calvert County Library Prince Frederick, thryland 20678 15/8 148 e
O
EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-06B BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS NUCLEAR POWER PLANT JNITS NO.1 & 2 DOCKET NOS:
50-317 & 318 Introduction By letter dated April 14, 1979, we transmitte.d I&E Bulletin No.79-06B to Baltimore Gas and Electric Company (BG&E or the licensee). This bulletin specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occurred at Three Mile Island, Unit No. 2 (TMI-2) on March 28, 1979. By letters dated April 26 and May 8,1979, BG&E provided their responses in conformance with the requirements of the bulletin for the Calvert Cliffs Nuclear Power Plant (CCNPP), Units No.1 & 2.
BG&E supplemented these responses by a letter dated June 5 and August 20, 1979 providing clari-fication and elaboration of certain of the items in response to our expressed concerns.
Our evaluation of these responses is given below.
Evaluation In this evaluation, the paragraph numbers correspond to the bulletin action items and to the licensee's response to each action item.
1.
BG&E states that all licensed operations' personel have been trained in the procedure changes made as a result of the TMI-2 accident.
In addi-tion, the requalification training lesson plans have been and will continue to be revised as a result of information from the accident.
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, BG&E has. reserved simulator time at the Combustion Engineering (CE) simulator to be totally devoted to accident scenarios including small-break simulations. The briefing in regards to the TMI-2 accident was presented by the NRC staff team consisting of Office of Inspection and Enforcement (I&E) and Operator Licensing Branch (OLB) representatives on April 19,1979. We find that the licensee has been responsive to the training requested by the reference bulletin.
2.
BG&E states that procedures governing the routine operation of pressurizer pressure control, chemical and volume control system (CVCS) and degasification system were reviewed and found to be adequate. The plant emergency procedure for a loss of coolant accident (LOCA) has been revised and expanded to emphasize the recognition and prevention of void formation in the reactor coolant system (RCS) and the enhancement of core cooling subsequent to void fomation should it occur.
2.a BG&E states that recognition of voiding due to steam formation can be by the determination of saturation conditions in the RCS using hot and cold temperature, in-core themocouples and pressurizer pressure instru-ments. The June 5,1979 submittal provides the ranges of the available instruments and describes the procedures to be used for forced and natural circulation modes of operation.
In addition, BG&E defines the
" core boiling-steam generator condensation" mode of core cooling and indicated that new procedures are available to provide guidance to the operators. We find the licensee's response in regards to the recogni-tion of possible void fomation during forced or natural cooling mode 13/8 d50
. of operation acceptable.
2.b To assist the operators in taking appropriate actions to prevent void formation, BG&E states that procedures have been reviewed and procedure changes, where necessary, have been impleniented. The revised procedures
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call for continudd operation of the high pressure safety injection (HPSI) pumps to provide make-up water and a source of system pressure.
Heat removal would be by forced or natural circulation or the reflux boiler mode previously mentioned.
In all three modes the heat sink is the steam generator. As an aid to the operator, curves displaying the 0
saturation line and a 50 F subcooling line have been added to the sub-ject procedure. We find that the licensee has adequately addressed the operator actions required to prevent void formation.
2, c In the August 20, 1979 response, the licensee states that the appropriate operator action required to enhance core cooling in the event core void-
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ing occurs is to restore pressurizer pressure and level and reinstate RCS cooling using the steam generators. Level is re-established using ECCS high pressure safety injection (HPSI) system pumps. Core cooling, provided by RCS flow through the steam generators, will nonnally be maintained by natural circulation according to the revised emergency procedures (See Section 6.c.).
BG&E states that the recovery of RCS pressure and continued core cooling will assure void collapse. We find that the licensee has adequately addressed this concern of the bulletin.
1 78 JS1 1
. 3.
In the design of the Calvert Cliffs Units, contairunent isolation signal (CIS) is automatically initiateu only on high containment pressure
($ 4.75psig). A Safety Injection Actuation Signal (SIAS) is initiated by high containment pressure (1 4.75 psig) or low pressurizer pressure
(>1573 psia). CIS is not automatically initiated by SIAS. BG&E states that a step has been added to the plant emergency procedure for loss of reactor coolarit requiring the containment penetrations of eight systems to be isolated manually from the control room upon receipt of a SIAS%
We find this procedure modification meets the intent of the bulletin requirements.
In addition, SG&E has initiated a design review to determine the feasiblity of automatic isolation of appropriate system penetrations upon a SIAS. The August 20, 1979 submittal indicates this design review is complete and seven isolation functions will be transferred from CIS to SIAS.
In addition, a design is being developed to assure a cooling water flow to the RCPs based on a logic-actuation system.
4.
BG&E states that even though the present Technical Specifications only require the presence of one licensed operator per unit in the control room, it is the management practice to augnent this staffing with an extra licensed Reactor Operator per unit when the unit is at power ~
operation.
This extra operator, having no further responsibility for reactivity control after the receipt of a plant trip, shall immediately direct his sole attention to the maintenance and/or restoration of feedwater flow via the main or auxiliary feed pumps as required by the situation.
In the event the manning level of the control room is altered such that the extra Reactor Operator is not available for each unf t,.a-person.specifically trained in the operation of the auxiliary feedwater system will be made available to assume these duties. The 13/8 052
5-operator assigned to this function will, at the time of a transient requiring such action, take immediate control of the main and auxiliary feedwater systems until the steam generator levels return to a stable condition. This extra operator will have no other concurrent respon-sibilities during such a transiente We find this response is in con-formance with the bulletin.
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5.
This bulletin item relates to the operation of the power operated relief valves (PORV's) on the pressurizer.
e 5.a BG&E states that plant operators may utilize the following control room indications to determine when a PORV is open. These indications con-sist of:
(1) a temperature indicator on the PORY common discharge header and (2) quench tank level, temperature and pressure indications.
We find such instrumentation satisfies the concern expressed in the bulletin and appropriate direction is provided by the emergency procedures.
5.b BG&E states that the plant emergency procedure for loss of reactor coolant was changed to identify specific instructions to be used in identifying an open PORY and to direct that the appropriate block valve be closed should one of the PORV's remain stuck open. The licensee's responses indicate that appropriate procedural control of a possible leaking PORV have been implemented.
6.
This bulletin item makes specific requests of licensees to ensure that
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procedures and tra'ning instructions pravent the overriding of engineered safety features during accident conditions.
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e 6
6.a BG&E states that procedures which require the override of Engineered Safety Featurc; (ESF) signals have been reviewed. They find that such overrides are deemed to be appropriate and not to cause any adverse effects on needed ESF needed to support core cooling. Of special interest is the, overriding of the CIS, once the containment pressure decays below the setpoint, in order to re-establish cooling flow to the RCP's and inhibit containment spray.
Since the component cooling water for the RCPs is isolated by a CIS, BG&E finds it important to reset CIS and re-establish cooling flow to the RCP's in order to prevent failure of the pump seals and the thrust bearing as explained in their June 5,1979 response. BG&E has initiated a design study to determine the feasibility of system modification to return component cooling flow to the RCP's in all post-accident situa-tions or to provide for the rapid restoration of the system following a non-seismic accident scenario. Recent communications indicate this design review is complete and some modifications will be made.
The licensee places special emphasis on securing the containment spray pumps, when not needed, to prevent damage to equipment within containment.
In recent comunications with BG&E, we learned that the procedure allows these pumps to be secured by overriding an automatic action only if the containment pressure is below 5 psig. In the Calvert Cliffs design, cantainment air recirculation units, redundant to the spray pumps, are available during accident conditions to handle containment cooling requirements.
1378 054 The licensee's responses and the above example indicate that procedural controls, preventing the overriding of automatic actions of engineered safety features have been initiated in accordance with the built.in.
6.b BG&E states that all of the criteria set forth in'I&E Bulletin 79-06B, Item 6.b have been incorporated into the appropriate procedures.
Although this adequately addresses the requirement of the bulletin, we are providing the following statement to clarify the intent of para-grsph 6.b.(2):
"After 50*F of subcooling has been achieved, termination of HPI operation prior to 20 minutes is only permissible if it has been determined that continued operation would result in an unsafe plant condition, e.g., pressure / temperature consider-ations for the vessel integrity."
6.c BG&E 's initial responses indicate that applicable emergency procedures have been revised to require continued operation of at least one RCP per lcop during the HPSI phase follcuing an accident.
BGSE stated that they would leave the RCP's running or will restart the pumps as long as the pump is pro-viding forced flow as indicated by :ontrol room indications. We find these statements responsive to the reouirements of Item 6.c of IE Bulletin 79-06B. However, the requirements of Action 6.c in IE Bulletin 79 06B were modified'by IE Bulletin 79-06C to trip the reactor coolant pun;ps, instead of
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keeping at least one pump running, after an initiation of high pressure injec-tion. This action is to,be taken by the licensee according to his August 20, 1979 response until the results of anlyses defined in IE Bulletin 79-06C are used to develop new guidelines for operator action. BG&E has evaluated its commitment 13/8 055
. in light of the requirements of IE Bulletin 79-06C and stated, in his August 20, 1979 letter, that the commitment as stated is consistent with the modified requirements. Our review of the licensee's responses is continuing pending the submittals of the long-term actions required by Bulletin 79-06C.
6.d The BG&E response states that the applicable emergency procedures have beenrevisedtofurtherminimizeoperatordependenceonpres:urizer
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level. We find that the licensee has adequately addres:ed this item of the bulletin.
7.
The licensee states that all safety related valve positions, positioning requirements and procedural controls, which ensure that the valves remain properly positioned, have been reviewed and are adequate to ensure proper operation of engineered safety features. The adminis-trative procedures for control of maintenance on safety related equipment were reviewed to specifically assure correct positioning of valves which were worked on or were used for isolation purposes. BG&E's letter of April 26, 1979 describes these procedural controls and pro-vides three actions taken to improve valve position control as a result of Bulletin 79-06B. We find the BG&E statements to be an acceptable response to this item of the bulletin.
8.
BG&E identified all systems designed to transfer potentially rt.Jioactive gases and liquids out of the primary containment and states that all of these systems are automatically isolated by a CIS and manually isolated upon receipt of a SIAS (Item 3).
In addition, the containment purge valves are closed upon detection of high 13/8
-)S6 radiation in the containment. To open an isolation valve follow-ing automatic closure by CIS, the initiating signal (high containment pressure) must haw cleared, the operator must manually reset the CIS signal and then position the valve hand switch to open. Containment isolation signals cannot be blocked before or after initiation and there are no process control systems which will automatically open a contain-ment isolation system valve. Once the containment isolation system is initiated the operator cannot override. Therefore, inadvertent trans-fer of radioactive fluids and/or gases will not occur during an incident.
!!e find that the licensee has acceptably addressed the bulletin concerns regcrding possible release of radioactive gases or liquids from the containment.
9.
Bulletin Item 9 relates to the safety-related system maintenance and test procedures.
9.a BG&E states that the administrative procedures specify that prior to removal of safety-related systems from service the redundant system will be verified operable by the shift supervisor. We find this con-cern of the bulletin has been properly addressed.
9.b The licensee states that when equipment (system or component) is returned to service after maintenance, it is the responsibility of the Senior Control Room Operator to verify the restoration of the equipment to service. Subsequent communication with BG&E has indicated that a 13/8 157 physical check of the equipment being returned to service is made as the operator removes the tags on this equipment. Upon completion of surveillance testing, the Shift Supervisor takes the ster, necessary to return the system to normal. We find this to be an adequate response to the bulletin request.
9.c BG&E states that whenever a safety-related system or component is taken out of service, the appropriate Technical Specification action state-ment is logged in the centrol room and shift supervisor's logs. Each succeeding shift must then enter a sunnary of these action statements as the first entry of each shift. This practice is carried on until the actions statement is cleared. We find this procedure ensures explicit notification of involved reactor operational personnel in regards to the status of safety-related systems and, therefore, is in conformance with this item of the bulletin.
10.
BG&E stated, in their August 20, 1979 response, that they have adopted a policy of reporting to the NRC within one hour any plant transient which causes the use of a plant Emergency Operating Procedure. We find this revision satisifes the concern of the bulletin, in that notification should be received when the plant enters a condition which is not controlied or expected.
The BG&E response requests more specific guidance on the installation of a " hot line" between the plant and Re'gion I headquarters. This line has been installed and agreement in regards to operation and testing has been worked out with I&E Region I.
Subsequent communication with the licensee indicates that this item has been satisfactorily resolved.
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. 11. BGSE states that the pmsent " degas" pmcedure using the volume contal tank is effec-tive during the mutine operations and could also be used in a post-accident situation.
They also indicate that two alternate methods are available and subsequent corm:unica-tions indicate that appmpriate procedures for these alternate methods am available.
The BGSE response describes the hydrogen reccrrbiners and the gas analyzing systems used to reduce and detemine the hydmgen concentration in the containment following an accident. They state that operatien of the reccrrbiners is governed by adminis-trative control and precedures.
BGSE's respense indicates an understanding of' the concern expmssed by this item of the bulletin. We therefore find BGSE's resocnse to this item acceptable.
CONCLUSIONS Based en our review of the inforntion previded by the licensee to date, we conclude that the licensee has correctly interpreted IE Bulletin No.79-06B.
'.'he actions taken demon-strate his understanding of the concerns arising frem the Three Mile Island, Unit No. 2 accident in mlation to their implications on his own operations, and pInvide added assurance for the protection of the public health and safety during plant operation.
This conclusion notwithstanding, it should be recognized that further actions may result frcm the staff's engoing review of cperating plants using nuclear steam supply systems designed by Ccnbustien Engineering. For example, the actions beirg taken for Iten 2 of IE Bulletin 79-06B regarding emergency procedures for a LOCA may require changes as a result of our generic review of precedures for C-E cperating plants. Additional changes may result firm the requirenents contained in NUREG-0578, e.g., the actions being taken for Item 5 of Bulletin 79-06B regarding the PCRV's. Our evaluations of these matters will be covered in cther reports.
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