ML13330A721
| ML13330A721 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 05/03/1979 |
| From: | James Drake SOUTHERN CALIFORNIA EDISON CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7906210348 | |
| Download: ML13330A721 (17) | |
Text
Southern California Edison Company P. 0. BOX 800 2244 WALNUT GROVE AVENUE J.H.DRAKE ROS CALIFORNIA 91770 TELEPHONE VICEPRE~pENT213-572-2258 VICE PPE510ENT May 3, 1979 37225 U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement' Region V Suite 202, Walnut Creek Plaza 1990 North California Boulevard Walnut Creek, California 94596 Attention: R. H. Engelken, Director Docket No. 50-206 San Onofre Unit 1
References:
(1) NRC (R. H. Engelken) letter to SCE (J. B. Moore) dated April 14, 1979, Docket No. 50-206.
(2) NRC (R. H. Engelken) letter to SCE (J.
B. Moore) dated April 18, 1979, Docket No. 50-206.
(3) SCE (J. T. Head, Jr.) letter to NRC (R. H. Engelken) dated April 19, 1979, Docket No. 50-206.
Dear Sir:
The following information is in reponse to References 1 and 2, which forwarded IE Bulletin 79-06A and Revision 1 thereto, concerning the recent Three Mile Island Incident.
In Reference (1),
you required action by us to immediately instruct our operating personnel that pressurizer level indications may not provide reli able transient or accident conditions.
Please be advised that we have so instructed operating personnel at San Onofre qnit 1 through the issuance of station memoranda dated April 12, 13 and 17, 1979.
Regarding Items 1-4 and 6-12 of IE Bulletin 79-06A and Revision 1 thereto, our responses are given below.
Our response to Item 5 concerning stationing of personnel to effect prompt manual initiation of auxiliary feedwater when re quired has been forwarded to your office under separate cover in Reference (3).
Our response to Item 13 concerning proposed changes to San Onofre Unit 1 tech nical specifications as a result of implementing Items 1-12 of 79-06A, and design changes necessary to effect long term resolution of these items will be forwarded to your office by May 23, 1979.
79062................................................................0
U.S. Nuclear Regulatory Commission Page 2 Item 1.
Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TI-2 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.
- a.
This review should be directed toward understanding:
(1) the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; (3) that the potential exists, under certain accident or transient conditions, to have a water level in the pressurizer simultaneously with the reactor vessel not full of water; and (4) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action.
- b.
Operational personnel should be instructed to:
(1) not override automatic action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 7a); and (2) not make operational decisions based solely on a single plant parameter indication when one or more confirmatory indications are available.
- c.
All licensed operators and plant management and supervisors with operational responsibilities -shall participate in this review and such participation shall be documented in plant records.
Response
L.a All NRC IE Bulletins concerning Three Mile Island which we receive are being distributed to San Onofre Unit 1 operating personnel for.
their review. In particular, IE Bulletins 79-05,79-05A, and en closures thereto were so distributed by station memoranda dated April 17 and 19, 1979, with explicit instructions that the accident circumstances and chronology described in these Bulletins be reviewed with the objective of understanding Items l.a(l) through l.a(4) above.
Regarding Item 1.a.(3),
it is noted that the instructions given to operating personnel concerning the potential for misleading pressur izer level indication have been based upon the events which occurred at Three Mile Island as described in IE Bulletin 79-05 and 79-05A.
San Onofre Unit 1 design basis accidents and transients are being reviewed to identify plant specific conditions wherein pressurizer level may riot be indicative of reactor vessel water level and to determine any required operator actions under such conditions.
The results of this review will be incorporated into operating instruc tions as appropriate.
~~
U.S. Nuclear Regulatory Commission Page 3 Response (Continued) l.b Standards governing reactor operating instructions, including those instructions which cover emergency conditions, are con tained in Station Order S-0-104.
This Station Order has been revised to incorporate the following requirements:
(1) The manual override of any emergency safety system shall not be initiated unless continued operation would result in unsafe plant conditions; and, (2) operational decisions shall not be based solely on a single plant parameter when more than one confirmatory indication is available.
Operating personnel are being instructed to review these require ments and implement them as appropriate when executing emergency instructions.
Emergency instructions will be reviewed to insure that the actions to be taken by operators as specified in those instructions are consistent with the revised Station Order.
l.c All licensed operators and plant management and supervisors with operational responsibilities assigned to and located at San Onofre Unit I have been instructed to participate in the review of IE Bulletins as described in l.a and l.b above. The participation of these individuals in this review is being documented in plant records.,
Item 2.
Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:
- a.
Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability.
- b.
Operator action required to prevent the formation of such voids.
- c.
Operator action required to enhance core cooling in the event such voids are formed (e.g. remote venting).
Response
2.a In Station memoranda dated April 12, 13 and 17, 1979, and in various station meetings, operating personnel have been instructed concerning
U.S. Nuclear Regulatory Commission Page 4 C
Response (Continued) the need to recognize the possibility of forming voids in the reactor coolant system during Unit I transient and accident conditions.
A positive indication that void formation has occurred in the reactor coolant system would be pressurizer pressure falling below the hot leg saturation pressure. To facilitate recognition of void formation, the control room operators have been furnished with a curve showing the saturation pressure which would corres pond to the hot leg temperature.
Regarding loss of coolant accidents in particular, void formation in the reactor coolant system is, in general, to be expected.
There are two exceptions wherein voiding would not be expected to occur during loss of coolant accidents.
These are 1) the loss of coolant is caused by a stuck open pressurizer relief valve which closes or is isolated before the reactor coolant system depres surizes to the hot leg saturation pressure, and 2) the break size is such that the reactor coolant system reaches an equilibrium pressure above hot leg saturation with safety injection flow equaling flow out the break. For these two cases, the operator could confirm general lack of voiding in the reactor coolant system by comparing pressurizer pressure with saturation pressure which corresponds to hot leg temperature. In any case, the San Onofre Unit I safety systems which are provided to mitigate the conse quences of loss of coolant accidents to within acceptable limits are designed to cope with attendant voiding.
Operating instructions for coping with transients and accidents are being reviewed in light of the circumstances and chronology of events associated with the Three Mile Island Incident in order to accomplish the following:
- 1)
Further determine the potential for forming voids in the primary system large enough to compromise core cooling capability, especially natural circulation capability;
- 2)
Identify means to enable operator recognition of possible void formation.
The results of this review will be incorporated into operating instructions and/or plant design, as appropriate.
2.b As indicated in the response to Item 2.a above, a review is being conducted to systematically determine the potential for 2.c void formation during San Onofre Unit I transients and accidents and the need for operator recognition of such voiding.
In con nection with this review, operator actions will be determined to:
U.S. Nuclear Regulatory Commission Page 5 Response (Continued)
- 1)
Prevent the formation of voids; and,
- 2)
Enhance core cooling or otherwise eliminate voids in the event they are formed.
The results of this review will be incorporated into San Onofre Unit 1 operating instructions and/or plant design, as appropriate.
Until this review can be accomplished and changes implemented, operating personnel will continue to be instructed concerning recognition of the possibility of forming voids and operator actions which would tend to prevent or eliminate their formation.
Such actions include: verify that reactor trip and safety injec tion have occurred; verify that reactor coolant temperature is not increasing; verify that pressurizer level is being maintained; maintain pressurizer pressure above the hot leg saturation pres sure by maintaining reactor coolant pressure by available means; (e.g. by charging pump and/or pressurizer heater operation); main tain reactor core cooling flow by continued safety injection opera tion or by operation of the safety injection recirculation system, and timely initiation of auxiliary feedwater to steam generators.
Operating personnel have been instructed concerning these actions through issuance of a station memorandum dated May 3, 1979.
Item 3.
For your facilities that use pressurizer water level coincident with pressurizer pressure for automatic initiation of safety injection into the reactor coolant system, trip the low pressurizer level setpoint bistables such that, when the pressurizer pressure reaches the low setpoint, safety injection would be initiated regardless of the pressurizer level. In addition, instruct operators to manually initiate safety injection when the pressurizer pressure indication reaches the actuation setpoint whether or not the level indication has dropped to the actuation setpoint.
Response
The San Onofre Unit 1 design provides for automatic initiation of safety injection upon receipt of 2 of 3 low pressurizer level signals coincident with 2 of 3 low pressurizer pressure signals. In response to NRC directives in Item 3 above, the following actions have been taken:
U.S. Nuclear Regulatory 96mission Page 6 Response (Continued)
- a.
The low pressurizer level bistables have been placed in trip such that, when the pressurizer pressure reaches the low set point, safety injection will be initiated irrespective of pressurizer level.
Provisions have been made for removal of the bistable trip during routine testing of the pressurizer pressure channels so as not to increase the possibility of spurious safety injection actuation during such testing.
Having taken the action noted we are now subjecting this plant design change to review by our internal review and audit com mittees as required by the administrative standards of the tech nical specifications incorporated into our Provisional Operating License.
This review is considered to be required by the provi sions of the Commission's Regulations, in particular 10 CFR 50.59.
- b.
Operating Instruction S-3-5.5 (Loss of Coolant) has been revised to require manual initiation of safety injection when pressurizer pressure falls below the actuation setpoint (1685 psig) on any two of the three pressurizer pressure channels and automatic initiation of safety injection has not occurred. This revision was accomplished on April 12, 1979.
Item 4.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to permit containment isolation, whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.
Response
The containment isolation design at San Onofre Unit 1 is such that the containment is automatically isolated upon receipt of a sphere high pressure signal (greater than 2 psig).
A review of station emergency procedures aid of the control circuitry of valves which receive this signal indicates that in all cases these valves may be closed by remote manual actuation from the control room at any time should such action be deemed necessary.
Item 5.
For facilities for which the auxiliary feedwater system is not auto matically initiated, prepare and implement immediately procedures which
U.S. Nuclear Regulatory Commission Page 7 require the stationing of an individual (with no other assigned concur rent duties and in direct and continuous ccimunication with the control room) to promptly initiate adequate auxiliary feedwater to the steam generator(s) for those transients or accidents, the consequences of which can be limited by such action.
Response
Our response to this item was forwarded under separate cover in Reference 3.
Item 6.
For your facilities, prepare and implement immediately procedures which:
- a.
Identify those plant indications (such as valve discharge piping temperature, valve position indication, or valve discharge relief tank temperature or pressure indication) which plant operators may utilize to determine that pressurizer power operated relief valve(s) are open, and
- b.
Direct the plant operators to manually close the power operated relief block valve(s) when reactor coolant system pressure is reduced to below the set point for normal automatic closure of the power operated relief valve(s) and the valve(s) remain stuck open.
Response
Operating Instruction S-3-5.4 identifies plant indications available to operators for use in determining that pressurizer power operated relief valves (PORV) are open.
Primary indications include valve position indication, PORV discharge line high temperature, and pres surizer relief tank high pressure, high demperature and high level.
Effective April 13, 1979, Operating Instruction S-3-5.4 was revised to direct operators to close the power operated block valve associated with a PORV for which there is indication that the PORV is open and reactor coolant system pressure is below the PORV actuation setpoint.
We are presently investigating the capability of the block valves to close under high flow conditions.
U.S. Nuclear Regulatory fmission Page 8 Item 7.
Review the action directed by the operating procedures and training instructions to ensure that:
- a.
Operators do not override automatic actions of engineered safety features, unless continued operation of engineered safety features will result in unsafe plant conditions. For example, if continued operation of engineered safety features would threaten reactor vessel integrity then the HPI should be secured (as noted in b(2) below).
- b.
Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:
(1)
Both low pressure injection (LPI) pumps are in operation and flowing for 20 minutes or longer; at a rate which would assure stable plant behavior; or (2) The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure.
If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated. The degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be limited by the pressure/temperature considerations for the vessel integrity.
- c.
Operating procedures currently, or are revised to, specify that in the event of HPI initiation with reactor coolant pumps (RCP) operating, at least one RCP shall remain operating for two loop plants and at least two RCP's shall remain operating for 3 or 4 loop plants as long as the pump(s) is providing forced flow.
- d.
Operators are provided additional irformation and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g. water inven tory in the reactor primary system.
Response
7.a Effective April 13, 1979, Station Order S-o-104, which establishes the standards for the operation and protection of reactor protec tion and control systems, was revised to explicitly prohibit
U.S. Nuclear Regulatory Commission Page 9 Response (Continued) manual override of emergency safety systems unless a careful con sideration and review of plant conditions indicate that continued operation of such systems would result in unsafe plant conditions.
This instruction has been reviewed with the operators and, by virtue of. its incorporation into the above referenced procedures, has been included in the operator training program.
7.b Effective April 16, 1979, Operating Instruction S-3-5.5 was revised to specifically preclude termination of safety in jection flow following automatic initiations except under the following conditions:
(1) It has been verified that a LOCA has occurred and indications are such that alignment is required of the safety injection recirculation system to take suction off the containment sump: or, (2) It has been determined that a Loss of Coolant has not occurred, the safety injection pumps have been in operation for at least 20 minutes and all hot and cold leg temperatures are and can be maintained at least 50 degrees below the saturation temperature for the existing reactor coolant system pressure. The degree of subcooling beyond 50 degrees F and the length of time beyond 20 minutes that the safety injection system is in operation shall be limited by the Technical Speci fications Figure 3.1.3b.
This instruction has been reviewed with the operators and is being incorporated into the operator training program.
7.c Presently, Operating Instruction S-3-5.5 (Loss of Coolant)
S-3-5.20 (Steam Line Break/Pipe Whip) and 5-3-5.31 (Steam Generator Tube Failure) restrict reactor coolant pump opera tion unless certain conditions are met following safety injec tion initiation.
These conditions are based upon the San Onofre Unit 1 safety analysis which does no,t address the consequences of reactor coolant pump operation for all cases subsequent to safety injection actuations.
Rather, the safety analysis demonstrates that accident consequences are within acceptable limits with stated restraints regarding reactor coolant pump operation.
Accordingly, we do not plan to alter operating instructions to specify continued operation of reactor coolant pumps in all cases following safety injection actuation.
7.d As stated earlier in this letter, station memoranda have been prepared and distributed to operating personnel concerning the
U.S. Nuclear Regulatory 9 mmission Page 10 Response (Continued) potential for unreliable reactor coolant system level indications using pressurizer level alone. In this regard, as noted above, the operators have been given general instructions to examine pressurizer pressure and reactor coolant temperature indications in addition to pressurizer level when evaluating plant conditions.
As stated in the response to Item l.a above, we are reviewing the matter of pressurizer level indications to identify conditions specific to San Onofre Unit 1 wherein unreliable indications may arise and to determine appropriate operator actions under such conditions. In addition, Station Order S-0-104, which defines standards for reactor-operations, has been revised to require that operational decisions not be based solely on a single plant parameter when more than one confirmatory indication is available.
This change was initiated on April 17, 1979. These changes are being incorporated into the operator training.program.
Item 8.
Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also review related procedures, such as those for maintenance, testing, plant and system startup, and supervisory periodic (e.g. daily/shift checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes.
Response
The station currently has procedures to control the positioning of valves critical to the functioning of engineered safety features.
These procedures include periodic surveillance requirements for operators, positive controls (i.e. locking of manual valves), and administrative controls for changing val~e positions. In light of the incident at Three Mile Island these procedures and the posi tioning requirements of all safety-related valves are under review to determine what changes, if any, may be required.
As stated in the response Item 10.b, Station Procedure S-A-107, Equipment Outage has been revised effective May 3, 1979 to require verification by operating personnel that safety related valves are properly returned to service following maintenance or test. Also, as stated in the response to Item 10.b, we plan to review plant administrative controls for all safety related systems to identify and make any changes to ensure proper verification of the operability of such systems when returned to service following maintenance or test.
U.S. Nuclear Regulatory Commission Page 11 Item 9.
Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
- a.
Whether interlocks exist to prevent transfer when high radiation exists, and
- b.
Whether such systems are isolated by the containment isolation signal.
- c.
The basis on which continued operability of the above features is assured.
Response
The above review has been completed. A table summarizing the applicable systems and the responses of their isolation valves is attached. As noted in the response to Item 4 above, to pre vent the reopening of the system isolation valves in the attached table when the interlocks and/or isolation signals are reset, the Operating Instruction S-3-5.5, Loss of Coolant, has been revised to require the operator to manually activate the containment iso lation signal when he determines that a loss of coolant has occurred.
This action is required prior to resetting safety injection and will block the reopening of these valves.
The basis for which continued operability of these features is as sured is provided by functional testing gf the safety injection system signal, the containment isolation signal and the isolation valves at frequencies specified by the Technical Specifications.
The high radiation interlock is tested weekly per station operating instructions.
Based upon our review of this matter and the revision made to Opera ting Instruction S-3-5.5, it is concluded that undesired pumping, venting, or other releases of radioactive gases and liquids from the primary containment will not occur inadvertently.
U.S. Nuclear Regulatory Commission Page 12 Response (Continued)
SYSTEMS DESIGNED TO TRANSFER POTENTIALLY RADIOACTIVE GASES AND LIQUIDS OUT OF PRIMARY CONTAINMENT ISOLATED BY HIGH ISOLATED BY SAFETY ESF RESET RADIATION CONTAINMENT INJECTION ISOLATION SYSTEM INTERLOCK ISOLATION INITIATION VALVES WILL Air Conditioning Yes, Note 1 Yes Yes Return to original positions Chemical and No No, Note 2 No Remain as is Volume Control Reactor Cycle No Yes, Note 3 No Outside-Return Sampling System to original position -
Inside Remain as is Radioactive Waste No Yes No Return to original Disposal System position Safety Injection No No, Note 4 No Remain as is System NOTES:
- 1.
The containment sphere atmosphere sample connections are not equipped with high radiation interlocks.
- 2.
The reactor coolant letdown and reactor coolant pump seal water return are isolated manually by remote means las part of the manual operator actions of the operating instruction for Loss of Coolant.
- 3.
The system outside isolation valves shut on the containment isolation signal.
The inside isolation valves are remote manually controlled by the operator.
- 4.
This system is pressurized post-LOCA to maintain recirculation cooling or safety injection.
As such they are considered an extension of the containment boundary and do not isolate on the containment isolation signal.
U.S. Nuclear Regulatory Commission Page 13 Item 10.
Review and modify as necessary your maintenance and test procedures to ensure that they require:
- a.
Verification, by test or inspection, of the operability of redundant safety-related systems prior to the removal of any safety-related system from service.
- b.
Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.
- c.
Explicit notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.
Response
10.a Station procedure S-0-111, Equipment Testing Before and After Maintenance, currently requires testing of specific safety-related equipment before the redundant piece of equipment is disabled.
We have reviewed the equipment identified in this procedure and veri fied that all redundant safety-related equipment identified by our Technical Specifications is required to be so tested.
10.b Station Procedure S-0-111, Equipment Testing Before and After Maintenance, requires testing the operability of redundant safety-related equipment, control valves, motor operated valves, power operated valves, and solenoid valves when they are returned to service following maintenance. In addition post maintenance testing is required for all valves and pumps identified in our ASME Section XI program for the inservice testing of pumps and valves. Post maintenance testing is also addressed in various maintenance and quality control inspection procedures.
General instructions governing removal of plant equipment from service for maintenance or repair and returning such equipment from service (i.e., clearance procedures) are contained in Division Order D-A-14, Work Authorizations. This order requires approval of the operator in charge to clear equip ment and requires tagging of cleared equipment.
Division Order D-A-ll establishes requirements for logging by control operators of clearances, including methods used to clear equipment, and clearance releases.
Station Procedure S-A-107, Equipment Outages, defines more specifically the responsibilities and procedures for preparing and approving requests to remove Unit 1 equipment from service that is important to safe, reliable operation of the plant.
U.S. Nuclear Regulatory Commission Page 14 Response (Continued)
We plan to review plant administrative controls governing main tenance and testing of all safety related systems and to identify and make such changes as may be necessary to ensure proper veri fication of the operability of such systems when returned to service following maintenance or test.
In addition, Station Procedure S-A-107, Equipment Outages, has been revised effective Pay 3, 1979 to require explicit verification by operating personnel that all safety related systems have been properly returned to service following maintenance and test.
10.c Station Procedures S-A-107, Equipment Outages, and S-0-111, Equip ment Tsting Before and After Maintenance, govern operator notifi cation when systems or components are removed and returned to service.
Effective April 23, 1979, these procedures have been modified to ex plicitly require notification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.
Item 11.
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a con trolled or expected condition of operation. Further, at that time an open continuous communication channel shall be established and maintained with NRC.
Response
Operating Procedures S-E-203 and S-VIII-1.4 have been reviewed and modi fied to assure that the NRC is notified within one hour and that following the initial notification, a continuous open communication channel with the NRC Region V IE Office will be maintained.
An additional telephone circuit will be installed at the on-site coordination center to specifically accom modate this channel. Installation of the additional circuit will be completed by May 18, 1979, and until the additional ckircuit is installed existing cir cuits would be utilized.
Item 12.
Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment.
00 U.S. Nuclear Regulatory Commission Page 15
Response
The post accident generation and control of hydrogen in containment and the primary system has been reviewed relative to the Three Mile Island Incident experience. We find there are two significant design differences between SONGS 1 and most other PWR's regarding hydrogen generation potential. Both of these differences would reduce the hydrogen generation following an accident at San Onofre Unit 1. First, the stainless steel fuel cladding in SONGS 1 affords major protection against hydrogen generations such that appreciable hydrogen from this source is not expected under conditions of the most severe design basis events. The other consideration is the fact that hydrazine is employed in the containment spray rather than NaOH to enhance iodine absorption.
The near neutral.pH of the spray water markedly decreases the hydrogen generation rates with susceptible metals.
In view of these considerations, it is our opinion that significant amounts of hydrogen gas will not be generated during transients or accidents at San Onofre Unit 1.
The modes for removing hydrogen from the reactor coolant system are:
- 1.
Hydrogen can be stripped from the reactor coolant to the pressurizer vapor space by pressurizer spray operation if the reactor coolant pump can be operated.
- 2.
Hydrogen in the pressurizer vapor space can be vented by power operated relief valves to the pressurizer relief tank.
- 3.
Hydrogen can be removed from the reactor coolant system by the letdown line and stripped in the volume control tank where it enters the waste gas system.
- 4.
In the event of a LOCA, hydrogen would vent with the steam to the containment.
If for some reason a non-condensible gas bubble becomes situated some where in the primary coolant system, there are many options for con tinued core cooling and removing the bubble,.
With a gas bubble located in the upper head several methods of core cooling are unaffected. The steam generator can be used to remove decay heat using reactor coolant pump forced flow or natural circu lation.
The safety injection system can be used to cool the core while venting through the pressurizer power operated relief valves.
Core cooling by any of these methods can proceed indefinitely if the primary coolant pressure is held constant.
If a lower system pres sure is desired, a controlled depressurization will allow the bubble to grow slowly until it uncovers the top of the hot legs.
U.S. Nuclear Regulatory Cdi.ission Page 16 Response (Continued)
This controlled depressurization can be performed in two ways:
- 1.
If the reactor coolant pumps can be operated, depressurization can be performed with a steam bubble in the pressurizer. De pressurization would be through the pressurizer power operated relief valves. Extra control is achieved with the pressurizer heaters and sprays if available. As the bubble grows to the top of the hot leg, small bubbles are carried through the system.
Degassing is done with the spray line and/or the Chemical and Volume Control System. The steam generator will carry away decay heat.
- 2.
If the reactor coolant pumps cannot be operated or their opera tion is undesirable, the pressurizer can be made water solid with the safety injection pumps running and the power operated relief valve and/or sample valve open.
Depressurization can be controlled by judicious use of the various valves, lines and pumps available in the safety injection system and by adjusting the pressurizer relief valves and/or sample valve.
As the bubble grows to the top of the hot leg, it slides across the hot leg and up into the steam generators.
As depressurization continues the gas bubbles grow in the steam generators and upper head but the core remains covered and cooled by safety injection water.
If there is enough gas, the pressurizer surge line would eventually be "uncovered".
Some of the gas would burp into the pressurizer and out the relief and/or sample valves.
This burping process would continue until the system were at the desired pressure. At that time this cooling mode could be continued or the system could be placed in an RHR mode.
It is noted that a gas bubble cannot be located in the steam generator with the reactor coolant pumps running. If a gas bubble forms in the steam generator during natural circulation, the re actor coolant pumps could be turned back on for degassing or safety injection flow could be initiated with the power operated relief valves open.
It is further noted that the gas bubbles cannot uncover the core in the above depressurization scheme because it will always tend to float to the top qf the system and it cannot compress water.
As noted previously, the hydrogen generation potential of SONGS 1 is significantly lower than most other PWR's particularly in the shorter term.
Eventual removal of hydrogen in containment can be accomplished by controlled venting.
Operating personnel have been instructed in regard to the above des cribed modes and procedures dealing with hydrogen generation and control.
Revisions to operating instructions to incorporate appropriate actions are presently being considered.
U. S. Nuclear Regulatory Commission Page 17 Item 13.
Proposed changes, as required, to those technical specifications which must be modified as a result of your implementing the above items and identify design changes necessary in order to effect long term resolu tions of these items.
Response
As stated earlier, the information requested in this item will be for warded to your office on or before May 23, 1979.
Many of the outstanding actions identified in the responses to items 1 -
12 above will require extensive reviews of plant design and operating pro cedures in conjunction with plant design basis accidents and transients.
As such, these actions will require support from our reactor supplier.
Therefore, we intend to base our schedule for completing these outstanding actions on communications with the reactor supplier and will include this schedule with our response to item 13.
Should you have any questions or require additional information, do not hesitate to contact me.
Sincerely, cc:
Director, Office of Inspection and Enforcement, Division of Reactor. Operations Inspection