ML19241B396
| ML19241B396 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/25/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | Stewart FLORIDA POWER CORP. |
| References | |
| FOIA-79-98 NUDOCS 7907160368 | |
| Download: ML19241B396 (22) | |
Text
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f UNITED STATES 3%
NUCLEAR REGULATORY COMMISSION 3
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WASHINGTON, D. C. 20555
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,j Docket No. 50-302 Mr. W. P. Stewart dg Director, Power Production 4X4 Florida Power Corporation
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P. O. Box 14042, Mail Stop C-4 y
St. Petersburg, Florida 33733
Dear Mr. Stewart:
We have completed a review of the information you have provided in letters dated April 9,1979, and April 12, 1979, in response to IE Bulletin 79-05A. provides an evaluation of your responses and discusses them with respect to their specificity, completeness, or the interpretation you have given to the require-ments of this Bulletin.
The evaluation also discusses areas of concern to the staff additional to those contained in the Bulletin.
In several places in Enclosure 1, we have identified additional action or infonnation that we require you to provide.
These items have been sumarized in Enclosure 2.
We request that responses to the items of Enclosure 2 be forwarded within 30 days of receipt of this letter.
Sincerely,
Enclosures:
1.
Evaluation Report 2.
Request for Additional Information r'gj 5520S3 v
Cce 729g 79071603h g'
UNITED STATES a[A
'i NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
%, -..... f EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-05A FLORIDA POWER CORPORATION CRYSTAL RIVER, UNIT NO. 3 00CKET NO. 50-302 Introduction By letter dated April 1,1979 we transmitted to the Florida Power Corporation (the licensee), IE Bulletin NO. 79-05.
This Bulletin specified actions to be taken by the licensee to preclude occurrence of an event similar to that which occurred at Three Mile Island, Unit No. 2 (TMI-2) on March 28, 1979.
This Bulletin was expanded and revised by IE Bulletin 79-05A (the Bulletin).
By letters dated April 9 and 12,1979, the licensee provided his responses in conformance with the requirements of the Bulletin.
Cur evaluation of these responses is given below.
In performing this evaluation we have utilized the additional clarification pro-vided in IE Bulletin 79-06A issued on April 14, 1979.
We have also evaluated the licensee's response related to those additional require-ments specified in t lletin 79-06A which are avolicable to Crystal River Unit No. 3.
552094 rm ET umu a
. EVALUATION Item 1 As a result of his review of the preliminary sequence of events of the incident at TMI-2 described in the IE Bulletin 79-05A, the licensee, ass'sted by Babcock and Wilcox Company, has identified the same six statements of concern outlined in the " Description of Circumstances" of the Bulletin.
He has also related these statements of concern to the Crystal River Unit No. 3.
We have reviewed his responses to these statements and find, in general, a satisfactory understanding of the sequence of events and their relationship to his facility.
It is not clear, however, that the licensee's review has been sufficiently broadbased with respect to the personnel in-volved in the review.
Therefore, to assure that all appropriate levels of the utility personnel have participated in this review, we will request a commitment that all licensed operators and plant management and supervisors with operational responsibilities par-ticipate in this review and such participation be promptly concluded and documented in plant records.
In response to Statement 1 of this item concerning the loss of feed-water, the licensee states procedures are being revised as necessary.
We believe licensees should specify those procedures being revised and the schedule for implementation.
552095
. In response to Statement 2 of Item 1 concerning the electromatic relief valve (ERV), the licensee has implemented procedural changes.
It is not clear, however, whether these procedures and instructions to the operators concerning the ERV, outline and emphasize the role of the block or isolation valve for the ERV.
We will request the licensee to 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 tem-perature, pressure, or level 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.
55299G Statement 3 of Item 1 relates to erroneous pressurizer level indica-tion. The licensee responded to this condition by referring to the response to Item 4.
In the licensee's response to Statement 4 in Item 1, we have a concern here as elsewhere throughout his response regarding the identification of the operating procedures which were reviewed and revised. We will request in all cases where reference is made to review, revision, or preparation of new procedures of training instructions, that the licensee provide the list of procedures by number and title of all procedures reviewed, revised, or prepared as it applies to specific concerns.
If additional aids are provided the operator (reactor coolant system pressure / temperature curves, for example), these need to be indicated as to the nature of the aid and the appropriate reference prccedure and training instruction.
Statement 5 of Item 1 relates to the possibility of pumping highly radioactive water from the relief valve discharge out of the con-tainment.
However, Item 9 of the Bulletin also requires the licensee to review operating modes and procedures to further assure that undesired pumping of radioactive liquids and gases out of the con-tainment will not occur inadvertently.
This is addressed in our evaluation of the response to Item 9 given below.
552037 In response to Statement 6 of Item 1, the applicant discusses revision to procedures that will be prepared in response to Bulletin Item 4 and which will preclude " premature" termination of reactor coolant pump operation.
Our evaluation of this matter is included in our evaluation of the licensee's response to Item 4 Although not specifically detailed in Item 1 of the Bulletin, we will request further information and commitments from the licensee concerning the review of operating modes and procedures to deal with significant amounts of hydrogen that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment.
Item 2 In response to Item 2 of the Bulletin, the licensee stated that he had reviewed transients which had occurred at Crystal River, Unit No. 3 to determine if any had elements similar to the chronology of events at Three Mile Island Unit No. 2 and Davis Besse Unit No.1.
Based on this review, the licensee submitted operating data on six transients, listed below, which contain similar elements of the chronology identified in IE Bulletin 79-05A.
552GS8 a.
Shutdown from Outside Control Room Test, April 16-23, 1977.
b.
Partial Loss of Power to ICS, April 21, 1977.
c.
Loss of Inverter "A"/ loss of Vital Bus "A",
Oct. 26, 1977.
d.
Excessive Cooldown Due to Stuck FW Block Valve, Jan, 6,1979.
e.
Turbine Bldg. Flooding / Loss of Feedwater, Jan, 17, 1979.
f.
Loss of Feedwater Flow to the "B" OTSG, Jan. 30, 1979.
These transients will, during the staff's on-going evaluation of the TMI-2 incident, be reviewed to determine whether further changes or modifications may be desirable to give added assurance that a TMI-2 accident will not be repeated.
It is further noted that two of these transients involve loss of power to the Integrated Control System (ICS) for which a procedure should be made available to the reactor operator. In connection with these transients we wi'11 recuest the licensee to inform us of any orocedures that existto guide the operator to respond to any cf these modes of failure in the IC3.
@'4033
. Item 3 We have reviewed the licensee's response and find it acceptable.
- However, we will request the licensee to identify the procedures reviewed and those revised as a result of his review and in addition we will request his schedule for implementation of these revised procedures.
Item 4 Item 4a The licensee response to this item states that abnormal and emergency procedures have been reviewed in order to ensure that operators do not override automatic action of ESF.
The response is not clear as to terminating the ESF.
The licensee will be requested to indicate presently anticipated criteria under which engineered safety features would be shutdown or restarted to prevent an unsafe plant condition.
In particular the licensee will be request to commit to Item 7a of Bulletin 79-06A which supplements Item 4a of Bulletin 79-05A and which adds the provision that operators do not override automatic actions of engineered safety features, unless continued operation of engineered safety features will result in unsafe plant conditions.
Item ab The licensee states that the procedures for termination of high pressure injection are being revised in accordance with IE Bulletin 79-05A.
55%100
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_ Item 4b of Bulletin 79-06A provides a revision and clarification of Item 4b of Bulletin 79-05A. We will request the licensee to review his procedures to be in accordance with IE Bulletin 79-06A.
Item 4c The licensee's response to this item states that operating procedures are being revised to specify that in the event of high pressure in-jection initiation with reactor coolant pumps (RCP) operating at least one RCP per loop shall remain operating.
We consider the licensee's response to this item to be acceptable.
However, we will request the licensee to expand his procedures in accordance with Item 7c of Bulletin 79-06A which specifies the operation of RCP as long as the pump (s) is providing forced flow and to identify the parameters that indicate forced flow.
Item 4d The licensee states that the operators have been provided additional information and instruction to not rely upon pressurizer level in-dication alone but to examine pressurizer pressure and other plant parameters.
It is not clear from this response whether the nature of the direction provided includes, for example, the pressure /
temperature relationship, the use of incore temperature readout,tave' at the steam generators, reactor outlet temperature and secondary side therredynamic conditions.
We will request that the licensee provide details of these parameters regarding the nature of the direction given to the operators in response to this item.
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Item 5 The licensee's response indicates that procedures providing valve line ups for engineered safety features have been reviewed and that valve positions have been verified against these procedures.
In our review we note that several valves appear to be missing from the list given in the response.
We will request the licensee to confirm that all appropriate valves (e.g., those valves on suction side of the auxiliary pump) are included in his review.
In addition, we will request he provide a list of the engineered safety features reviewed and the identi-fication of the appropriate operating, emergency, maintenance and testing procedures together with the schedule implementation.
Item 6 In response to this item, the licensee has provided an extensive list of all valve and identifies the valves that are required for core cooling and the valves that receive isolation signals.
He is reviewing this list to identify which additional valves will be modified to receive isolation signals upon automatic HPI actuation.
In order to assure that the licensee's review will provide the necessary safeguards we will request the licensee to provide his implementation schedule and to justify not isolating (whether manual or automatic) those lines whose isolation does not degrade core cooling capability upon automatic safety injection and to justify all lines required for core cooling.
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, In addition, the licensee will be requested to review and identify operating procedures regarding containment isolation.
Iten 7 We have reviewed the licensee's response concerning positive position controls for manual valves mud manually-operated motor driven valves, and find that he has adequately addressed the item as stated in the Bulletin.
However, we will request that he review his maintenance and surveillance procedures that assure the operability of the two pneumatically operated valves in the auxiliary feedwater system, and that he provided information concerning indications of the positions of these valves.
We will also request thct the lic:nsec indicate his comri;itment to verify the correct position E: to the ncnucl valvas addressed in this item which are locked, sealed or otherwise secured.
Item 8 The licensee responded to this item by identifying the independent flow paths for the auxiliary feedwater flow paths. The special operating procedure SP349 " Emergency Feedwater System Operability Demonstration" has been revised to meet the requirement of this item. A technical sp'cification change request will be submitted to include the action statement of this item into the technical specification.
Until a license. amendment is issued, these actions will be administratively enfo rced.
We find the licensee's response acceptable.
b52103
. Item 9 The licensee in responding to this item, identified eleven items per-taining to system capable of pumping radioactive gases or liquids from containment after reviewing all systems and components that are part of reactor building isolation including cooling systems.
Two systems designed to transfer potentially radioactive gases and liquids are the reactor building purge and the reactor building sump.
These systems are automatically actuated from actions performed by the engineered safeguard system.
The reactor building purge is isolated by a preset radiation level and purge is activated by operator procedure.
The controls for reactor building sump are being modified to be interlocked with the actuation of high pressure injection (HPI) signal to assure against inadvertent pumping of the sump during HPI.
The licensee is also modifying the control to the reactor drain tank pump :nd all waste gas connections to shut down by the high pressure injection signal.
Although we find this item acceptable the licensee will be requested to furnish the following infomation:
SU'e.d.04
. 1.
A schedule for the implementation of containment isolation by HPI initiation.
2.
A basis for reactivation of items isolated by the safety injection after safety injection is terminated.
3.
Identify the procedures that are affected by these modi-fications.
4.
The time period required to implement the short term instruction described under this item.
This item was.also modified by IE Bulletin 79-06A. ~ The licensee will be requested to complete his response by responding to Item 9c of IE Bulletin 79-06A.
The licensee will be requested to identify the operation modes and procedures which were reviewed.
The subject of the containment isolation valves being opened to allow purging during operation is presently under staff review.
In our letter of November 29, 1978, we requested that the licensee provide a justification for continued purging at his facility and to limit purging to an absolute minimum, not to exceed 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year, pending the NRC staff review of his justification.
The licensee provided a justification for purging in his letter of January 10, 1979 and his response is under review.
552105
- Item 10a The licensee's verification of the operability of redundant safety-related systems include either the performance of the appropriate surveillance on the redundant system or the confirmation that the technical speci-fication's (TS) interval for performance prior to return to service would not be exceeded.
The TS provides a list of safety features for which this requirement is acceptable.
The auxiliary feedwater system is not included in the list, but none-theless it must be regarded as a safety-related system within the con-text of the term " safety-related" as used in the Bulletin.
We will request the licensee to review the operation of Crystal River to deter-mine if other systems should be included in the scope of his review.
Item 10b We find the licensee's response concerning the verification of opera-bility of safety-related systems following maintenance is acceptable provided that there is verification that the system has been returned to its normal operating configuration.
We will request the licensee to confirm this verification, and will request that he state what, if any, confirming checks exist in the verification process.
5521CG o
. Item 10c We consider that the licensee's procedure concerning maintenance on safety-related systems will keep the Shift Supervisor continually informed of the status of these systems.
However, we will request that the licensee ensure that all other reactor operating personnel (for example, control room operators) are explicitly notified when-ever a safety-related system is removed from or returned to service.
Item 11 The licensee states that all operating personnel have had training and have been made aware of the extreme seriousness and consequences of the TMI-2 incident.
He further states that maintenance personnel do not change valve positions at Crystal River Unit No. 3 and have been instructed that they do not have this authority.
Item 12 The response from the licensee outlines the procedural controls that have been established for NRC notification of serious events.
We note that although reference is made by the licensee to the require-ments established in 10 CFR 20, no mention is made specifically of following the guidance of Regulatory Guide 1.16, " Reporting of Operating Information - Appendix A Technical Specifications," and Regulatory Guide 1.101, " Emergency Planning for Nuclear Power Plants."
We will request that the licensee clarify his response as to the in-corporation of the guidance provided by R. G.
1.15 and 1.101 8e 51.(.?
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. However, we believe additional emphasis should be placed by the licensee on the concern addressed in the Bulletin regarding very early notification of the NRC of serious events.
We will request that the licensee prepare and implement immediately necessary actions and procedures to assure that an open, continuous communication channel can be established and maintained with NRC and that the NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
_ CONCLUSION Based on our review of the information provided by the licensee to date, we conclude that, while we have identified certain areas wnere additional information or action is needed, the licensee has correctly interpreted IE Bulletin No.79-05A.
The actions taken demonstrate his understanding of the salient concerns arising from the Three Mile Island incident in reviewing these implications on his own operations, and provided added assurance for the protection of the public health and safety during plant operation.
Technical Specifications should be reviewed to determine necessary changes because of the requirements of the Bulletin.
We will request that the licensee propose changes, as required, to those technical specifications which must be mcdified as a result of implementing any revised procedures resulting from Bulletin 79-05A.
552108
ENCLOSURE 2 _
ADDITIONAL ACTIONS REQUESTED OF THE LICENSEE 1.
Ensure that all appropriate levels of utility personnel have parti-cipated in the review required by Item 1 of IE Bulletin 79-05A.
.All licensed operators, plant managers, and supervisors with opera-tional responsibilities should participate in this review, and such participation should be promptly concluded and documented in plant records.
2.
Prepare, and implement imediately, procedures which:
a.
Identify those plant indications (such as valve dis-charge piping temperature, valve position indication, or valve discharge relief tank temperature, pressure, or level indication) which plant operators may utilize to detennine 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.
CSUI' 552106' '
. 3.
In all cases where reference has been made to a review or revision of existing procedures, or a preparation of new procedures, provide a list of the procedures by number, title and identified to specific concerns.
If additional aids are provided to the operator (e.g., reactor coolant system pressure / temperature curves), these should be fully describes, and the appropriate procedures supporting the use of these aids should be referenced.
4.
Review operating modes and procedures that deal with significant amounts of hydrogen as that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment.
5.
Identify any procedures that exist to guide the operator. response to any of the various modes of failure in the Integrated Control System (ICS).
6.
In response to Item 4a and 5, provide a list of all procedures which you have reviewed including appropriate emergency procedures and identify those that were revised as a result of this review.
Provide a list of the engineered safety features included in your review.
552110
. 7.
Indicate presently anticipated criteria under which engineered safety features would be shutdown or restarted to prevent an unsafe plant condition.
In particular commit to Item 7a of Bulletin 79-06A which supplements Item 4A of Bulletin 79-05A and which adds the provision that operators do not override automatic actions of engineered safety features, unless continue'd operation of engineered safety features will result in unsafe plant conditions.
8.
In regard to terminating high pressure injection the review of the procedures are to include provision of Item 7b of Bulletin 79-06A.
9.
Expand your procedures to include provision of 7c of Bulletin 79-06A which specifies that operation of an RCP in each loop con-tinues as long as the pumps are providing forced flow and to identify the parameters t.'1at indicate " forced flow."
10.
Provide details of other parameters (e.g., pressurizer pressure /
temperature, tave, etc.) the operator will use in support of pressurizer-1evel indication.
11 Confirm that all appropriate valves (i.e., those valves on suction side of the auxiliary pump) are included in your review.
Provide a list of the engineered safety features reviewed for Item 5 and the identi'ffcation of tha appropfiate operatinc,' emergency maintenance and testing procedures together with the schedule implementation.
55Z111
' 12 Provide the implementation schedule and justify not isolating (whether manual or automatic) on those lines whose isola *_ ion does not degrade core cooling capability upon automatic safety injection and justify all lines required for core cooling.
In addition.
review and identify operating procedures regarding containment isolation.
13.
raview the maintenance and surveillance procedures that assure the operability of tae two pneumatically operated valves in the auxiliary feedwater system, and provide information concerning indications of the positions of these valves.
14.
Furnish the following information concerning with reactor building isolation:
a.
A schedule for the implementation of containment isolation by HPI initiation.
b.
A basis for reactivation of items isolated by the safety injection after safety injection is terminated.
Identify the procedures that are affected by these c.
modi fications.
d.
The time period required to implement the short
' bN
- term instruction described under this item.
U
. e.
Respond to Item 9c of IE Bulletin 79-06A and identify the operating modes and procedures which were reviewed.
15.
Briefly describe the procedures to be used to verify operability prior to and following removal of safety-related systems including the emergency feedwater system from service for maintenance or testing.
16.
Provide a list of safety features which were reviewed under Item 10a.
The auxiliary feedwater system is not included in the list, but nonetheless it must be regarded as a safety related system within the context of the term safety related as osed in the bulletin.
17.
Confirm the verification of operability of safety-related systems and what confirming checks exists in the verification process.
18.
Ensure that all other reactor operating personnel (for example Control Room Operators) are explicitly nr tified whenever a safety-related system is removed frcm or returned to service.
19.
In regard to NRC notification clarify that R. G.1.16 and 1.101 are incorporated in the procedures as guidance.
L,52113
. 20.
Prepare and implement immediately necessary actions and pro-cedures to assure that an open, continuous communication channel can be established and maintained with NRC and that the NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.
21.
Propose changes to those technical specifications which must be modified as a result of implementing any revised procedures resulting from Bulletin 79-05A.
22.
With respect to Item 7 indicate the frequency of check of the correct positioning of manual valve addressed in this item which are locied sealed or otherwise secured.
L S.', 1 1 4
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