ML19260C575
| ML19260C575 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/21/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Whitmer C GEORGIA POWER CO. |
| References | |
| IEB-79-08, IEB-79-8, NUDOCS 8001080035 | |
| Download: ML19260C575 (15) | |
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December 21, 1979 Docket Nos. 50-321 and 50-366 i
Mr. Cnarles F. Whit er Vice Presiden: - Engineering Georgia Pcwer Company P. 0. Box 4545 Atlanta, Georgia 30302 Dear ".r.
Whitmer:
SUBJ ECT:
LC STAFF EV'LUATION OF GECRGIA POWER COMPANY RESPCNSES TO IE EULLET:N 79-08 FOR HATCH NUCLEAR PLANT UNITS NOS. 1 AND 2 We have ccmaleted cur review of the information that you provided in your letters dated April 25 and May 9,1979 in response to IE Bulletin 79-08 for the E. I. Hatch Nuclear Plants.
We have also completed our review of the supplenental information that you provided in your letters of Augus: 10, 1979.
We have concluded that you have taken the appropriate actions to meet the requirements of each of the eleven action items identified in IE Bulletin 79-08. A copy cf cur evaluation is enclosed.
As you know, NRC staff review of the Three Mile Island, Unit 2 (TMI-2) accident is cont nuing and other corrective actions may be required at a d
later cate.
For exaccle, the Bulletins and Orders Task Force is conduct-ing a generic revies of operating boiling water reactor plants. Specific requirenents for ycur facility that result from this and other TMI-2 investigations will be addressed to you in separate correspondence.
Si ncerely,
7 4f C-Tn,. 6.%::omas%.%ppofito, Chief Operating Reactors Branch #3 Division of Operating Reactors Enci csure:
NRC Staff Evaluaticn cc w/ enclosure:
See next page 1701 219 8001080 @6
Mr. Charles F. Whitmer Georgia Power Company cc:
G. F. Trowbridge, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N. W.
Washington, D. C.
20036 Ruble A. Thomas Vice President P. O. Box 2625 Southern Services, Inc.
Bint.ingham, Alabama 35202 Ozen Batum P. O. Ocx 2 2:
Southern Services, Inc.
Birmingham, Alabama 35202 Mr. William Widner Georgia Power Comoany Power Generation Department P. O. Box 4545 Atlanta, Georgia 30302 Mr. L. T. Gucwa Georgia Power Company Engineering Depart. gent P. O. Box 4545 Atlanta, Georgia 30302 Appling County Public Library Parker Street Baxley, Georgia 31413 Mr. R. F. Rogers U. S. Nuclear Regulatory Commission P. O. Box 710 Baxley, Georgia 31513 1701 2?0
EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-08 GEORGIA POWER COMPANY EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 DOCKET N05. 50-321 AND 50-366 1701 221
Introduction By letter dated April 14, 1979, we transmitted Office of Inspection and Enforcement (IE)Bulletin 79-08 to Georgia Power Company (GPC or the licensee).
IE 3ulletin 79-C8 specified actions to be taken by the licensee to avoid occurrence of an event simi'ar to that which occurred at Three Mile Island, Unit 2 (TMI-2) on March 28, 1979.
By letter dated April 25, 1979, GPC Orovided initial responses to Action Items 1 tnrougn 10 of IE Bulletin 79-08 for the Hatch Nuclear Plant, Units 1 and 2 (HNP 1 & 2).
GPC supplemented this rss;cnse by a letter dated May 9,1979 to provide the response to Action I en 11 of IE Eulletin 7C-08, and to revise ar.d clarify Items 1 through 10.
Tne NRC staff review of the GPC responses led to the issuance of requests for additional informaticn regarding the GPC resocases to certain action items of IE Sulletin 79-08.
These requests were contair.ed in a letter dated July 20, 1979.
By letter dated August 10, 1979, GPC responded to the staf f's requests f or additional information.
T.ne GPC responses to IE Bulletin 79-08 provided the basis for our evaluation
- esented below
In acdition, the actions taken by the licensee ic response to the bulletin requirements and subsequent NRC recuests were verified through inspections by IE inspectors.
Evaluation Each of the 11 action items requested by IE Bulletin 79-08 is repeated below fM' owed by our criteria for evaluating the response, a summary of the licensee's response and our evaluation of tre response.
~.
Review the cescription of circumstances cescribed in Enclosure 1 of IE Bulletin 79-05 anc the preliminary chronology of the TMI-2 Marcn 28,1979 accicent incluced in Enclosure 1 to IE Sulletin 79-05;.
a.
Inis res'ew should be directed toward anderstanding:
(1) tne extrere seriousness and consequences of the simultaneous bicckin; of ooth trains of a safety system at the Three Mile 1701 222
2 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; and (3) 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 Sa of this bulletin); 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 o,ne"Honal responsibilities shall particioate in this review and such participation shall be documented in plant records.
The licensee's response was evaluated to determine that (1) the scoDe of review was adequate, (2) operational personnel were properly instructed and (3) personnel participation in the review was documented in plant records.
The licensee's response dated April 25, 1979 indicated that the required training had been completed and documented on training data sheets, except for tnree licensed personnel who were not available at the time the training was conducted.
The licensee's revised response dated May 9, 1979 indicated that the required training had been completed and documented for all licensed personnel.
Wc conclude that the licensee's scope of review, instructions to operating personnel and documented participation satisfies the intent of IE Bulletin 79-08, Item 1.
2.
Review the containment 1 solation initiation design and procedures, and prepare and implement all changes necessary to initiate 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.
1701 223
3 The licensee's response was evaluated to verify that containment isolation initiation design and procedures had been reviewed to assure that (1) manual or automatic initiation of containment isolation occurs on automatic initia-tion o' safety injection and (2) all lines (including those designed to transfer radioactive gases or liquids) whose isolation does not degrade cooling capability or needed safety features were addressed.
The licensee's May 9,1979 response identified nitrogen
'arting make-up lines in Unit 1 anc two-inch purge bypass lines in Units 1 and 2 that do not aut:matically isolate upon initiation of safety injection.
The Unit 1 containment nitrogen inerting system does not automatically isolate so as to enable inerting the primary containment atmosphere with nitrogen during post-accident conditions.
In a supplemental response dated August 10, 1979, the licensee committed to modify the design of the two-inch purge bypass lines in both units to provide for automatic isolation upon safety injection.
We conclude that the licensee's review of containment isolation initiation design and procedures satisfy the intent of IE Bulletin 79-08, Item 2.
3.
Descrice the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e.g., RCIC) that are used wnen the main feedwater system is not operable.
For any manual action necessary, describe in summary form the procedure by which this action is taken in a timely sanse.
The licensee's response was reviewed to assure that (1) it described the automatic and manual actions necessary for the proper functioning of the auxiliary heat removal systems when the main feedwater system is not operable and (2) the procedures for any necessary manual actions were described in summary form.
The licensee's response dated May 9,1979 described the automatic actions recuirec when the main feedwater system is tot operable.
The reactor would automatically scram wnen the reactor sessei water level decreased to
+12h inches.
Should the level continue to cecrease to -38 inches, the high 1701 224
l>
4 l
i pressure coolant injection (HPCI) system and the reactor core isolation i
cooling (RCIC) system will initiate and inject into the reactor vessel.
We acknowledge the capability of these systems to provide the required heat removal action.
The operator can manually secure the HPCI system when reactor vessel level is confirmed to be above the low level scram point /+12 inches).
When reactor water level is restored and stabilized at +37 inches (normal i
level), the operator can secure the RCIC system.
i We conclude that the licensee's procedural summary of automatic / manual actions l
necessary for the proper functioning of auxiliary heat removal systems used when the mai-f::d.;;ter system is inoperable satisfies the intent of IE l
Bulletin 79-08, Item 3.
4 Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems.
Describe other redundant instrumentation which the operator might have to give the same informa-tion regarding plant status.
Instruct operators to utilize other avaiiable information to initiate safety systems.
The licensee's response was evaluated to determine that (1) all uses and types of vessel level
- indication for both automatic and manual initiation of safety systems were addressed, (2) it aadressed other instrumentation available to the operator to determine changes in reactor coolant inventory and (3) opera-tors were instructed to utilize other available information to initiate safety systems.
The licensee's May 9, 1979 response listed the various control room instruments used for monitoring and recording vessel level.
The indicated ranges of these instruments vary from 200 to 900 inches as measured from the bottom head drain.
In addition to the control room instruments, the response list.d several local instruments located in the reactor building, most of wnich dr. not require electrical power to operate.
The indicated ranges of these instruments vary from 200 to 577 incnes as measured ' rom the bottom head drain.
Manual initiation of safety systems can be based upon information from 19 :eparate indicators, 15 of which on Unit 1 and 14 of which on Unit 2 do not 1701 225
5 require an external power source.
Additional instrumentation which the operator can use to determine changes in reactor coolant inventory was addressed in a supplemental response dated August 10, 1979.
Licensed personnel were instructed in the use of all available instrumentation.
We conclude that the licensee's description of the uses and types of reactor vessel level / inventory instrumentation and instructions to operators regarding the use of this information satisfies the intent of IE Bulletin 79-08, Item 4.
5.
Review the actions directed by the operating procedures and training instructions to ensure that:
3.
Operators do not override automatic actions of engineered safety features, unless continued operation of engineered i
safety features will result in unsafe plant conditions (e.g.,
vessel integrity).
b.
Operatcrs are provided additional information and instructions to not rely upon vessel level indication alone for manual actions, but to also examine other plant parameter indications in evalating plant conditions.
The licensee's, response was evaluated to determine that (1) it addressed the matter of operators inproperly overriding the automatic actions of engineered safety features, (2) it addressed providing operators with additional informa-tion and instructions to not rely upon vessel level indication alone for manual actions and (3) that the review included operating procedures and training instructions.
The licensee in its May 9, 1979 response stated that HNP operating procedures caution the plant operators to ensure that the water level as well as the other plant parameters are normal before changing the system from its automatic function.
The supplemental response dated August 10, 1979 stated that the operators are instructed to not rely upon vessel level indication alone for manual action.
These instructions are provided by the training department and documented attendance is required.
1701 226
6 We conclude that the licensee's review of operating procedures and training instructions satisfies the intent of IE Bulletin 79-08, Item 5.
6.
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 start-up, 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.
The licensee's resconse was evaluated to assure that (1) safety-related valve positioning requirements were reviewed for correctness, (2) safety-related valves were verified to be in the correct position and (3) positive controls were in existence to maintain proper valve position during normal operation as well as during surveillance testing and maintenance.
The licensee's response dated May 9,1979 cescrit ed the review of safety-related valve positioning requirements.
Complete valve line-up checks are performed for safety-related valves prior to startup from an extended outage.
(Both plants were in cold shutdown when the initial response to the bulletin was prepared.) Administrative controls governing normal operation, surveillance testing, and maintenance were described.
The supplemental response dated August 10, 1979 confirmed that valve position and locked valve status are documented on data sheets.
We conclude that the licensee's review of safety-related valve positioning requirements, valve positions and positive controls to maintain proper valve positions satisfies the intent of IE Bulletin 79-08, Item 6.
7.
Review your operating modes and procedures for all systems cesigned to transfer potentially radioactive gases and liquids out cf the primary containment to assure that undesired pumping, venting or other release of radiaactise liquics 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 sucn systems and indicate:
1701 227
7 a.
Whether interlocks exist to prevent transfer when high radiation indication 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.
The licensee's response was evaluated to determine that (1) it addressed all systems designed to transfer potentially radioactive gases and liquids out of primary containment, (2) inadvertent releases do not occur on resetting engineered safety features instrumentation, (3) it addressed the existence of interlocks, (4) the systems are isolated on the containment isolation signal, (5) the basis for continued operability of the features was addressed and (6) a review of the procedures was performed.
In the May 9,1979 response, the licensee reported that potentially radioactive gases are transferred from containment through the drywell vent a
and purge system.
For Unit 1, the valves are part of the contain.nent atmos-pheric dilution (CAD) system and are opened by procedure under continuous operator supervision.
The operator is required to close the valves upon receipt of an isolation signal.
On Unit 2, the normal vent valves are used and they will automatically shut upon receipt of a reactor building or refueling area high radiation signal, or a containment isolation signal.
In the supplemental response dated August 10, 1979, the licensee committed to install plant modifications to isolate the valves that could inadvertently transfer radioactive gases from the containment upon reset of an isolation signal.
This modification is to be completed during the first outage of sufficient duration subsequent to receipt of engineering and materials.
In the May 9, 1979 response, the licensee reported that the radwaste sump subsystem is provided to transfer potentially radioactive liquids out of the
- imary containment.
The pump discharge valves are automatically closed by the containment isolation signals; however, they could re-open upon isolation 1701 228
8 signal reset, thus setting the stage for an inadvertent release.
Adminis-trative controls are presently in place to require an operator to close these valves prior to resetting.
In addition, an additional reset for these radsaste isolation valves will be installed at the next outage of sufficient duration subsequent to receipt of engineering and materials.
The additional reset will preclude an inadvertent release path from being established.
The annunciatcr response procedure requires the operator to manually close these valves on reactor building high radiation or refueling area high radiation (as well as containment isclation).
We conclude that the licansee's review of systems designed to transfer radioactive gases and liquids out of primary containment to assure that undesired pumping, venting, or other release of radioactive liquids and gases will not occur satisfies the intent of IE Sulletin 79-08, Item 7.
8.
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 safety-related systems when they are returned to service following maintenance or testing.
c.
Explicit nctification of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.
The licensee's response was evaluated to determine that operability of redundant safety-related systems is verified prior to the removal of any safety-related system frcm service. Where operability verification appeared only to rely on previous surveillance testing within Technical Specification intervals, ne asked tnat operability be further verified by at least a visual check of the system status to the extent practicable, prior to removing the redundant equipment from service.
The response was also evaluated to assure provisions were adequate to verify operability of safety-related systems when they are returned to service following maintenance or testing.
We also 1701 229
d l
l checked to see that all involved reactor operational personnel in the onccming shift are explicitly notified during shift turnover about the status of systems removed from or returned to service since their previous shif t.
The licensee's response dated May 9, 1979 indicated that operability of redundant safety-related systems was verified by inspection of clearance sheets and tags, verbal communications, and visual examination of the system inoperability status coard.
Some revisions to maintenance and test procedu-es were found to be necesary to assure provisiens were adequate to verify operability of safety-related systems when they are returned to service followinc maintenance or testing.
The licensee committed to con.plete the necessary procedural modifications prior to startup of the applicable unit.
The supplemental response dated August 10, 1979 described tne methods used to assure that all involved reactor operational personnel in the oncoming shift are explicitly notified during shift turnover about the status of systems removed from or returned to service since their previous shift.
We conclude that the licensee's review and codification of maintenance, test and administrative procedures to assure the availability of safety-relatec systems and operational personnel knowledge of system status satisfies the intent of IE Bulletin 79-08, Item 8.
9.
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 controlled or expected condition of operation.
Further, at that time an open continuous communication channel shall be established and maintained with NRC.
Tne licensee's response was evaluated to determine that (1) prompt reporting procedures required or were to be modified to require that the NRC is notified within one hour of tne time the reactor is nct in a controlled or expected condition of coeratic, and (2) procedures required or were to be modified to re Jre the establ:" ent and maintenance o' an open :cntinuous communication narnel with the NRC 'ollowing sucn events.
1701 230
10 The licensee reported in its May 9, 1979 response that the Technical Specifications and plant procedures required NRC notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
However, a commitment was made to modify the plant emergency procedures by July 1,1979 to require plant personnel to establish and maintain an open continuous communication channel with the NRC within one hour after deter-mining that an emergency condition exists.
The supplemental response dated August 10, 1979 indicated that the communication channel would be used where it is apparent that immediate NRC attention is necessary.
The licensee confirmed its intent to notify the NRC within one hour whenever the reactor is operating in an uncontrolled or unexpected condition by telephone on November 9.1974 We conclude that the licensee's response satisfies the intent of IE Bulletin l
79-08, Item 9.
10.
Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be gerierated during a transient or other accident that would either remain inside the primary system or be released to the containment.
The licensee's response was evaluated to determine if it described the means or systems available to remove hydrogen from the primary system as well as the treatment and control of hydrogen in the containment.
The licensee in its May 9,1979 response stated that it reviewed plant procedures which deal with significant munts 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.
Several discrepancies were identified and corrected in the post-accident venting procedure for HNP-2 which uses two 100 percent capacity hydrogen recombiner systems designed to control and recombine the hydrogen buildup within containment.
The HNP-1 containment is inerted with nitrogen and would employ the containment nitr: gen inerting system to contrcl the post-accident hydrogen bu ldup within i
containment.
Both units are equipped with the normal vent and relief lines to control hydrogen buildup in the primary system.
1701 231
11 We conclude that the licensee's response satisfies the intent of IE Bulletin 79-08, Item 10.
11.
Propose changes, as required, to those technical specifications which must ce modified as a result of your implementing the items above.
The licensee's response was evaluated to determine that a review of the Technical Specifications had been made to determine if any changes were required as a result of implementing Items 1 though 10 of IE Bulletin 79-03.
Tne licensee reported in its letter dated May 9,1979 that its review has s,cun that no changes to the Technical Spec'fications are required.
The licensee stated in subsequent telepnone conversations, that should modifica-tions to the Technical Specifications be recuired, they will be proposed in a timely manner.
We conclude that the licensee's response satisfies the intent of IE Bulletin 79-08, Item 11.
Conclusion Eased on our review of the informaticr. provided by the licensee to date, we c:nclude that the licensee has correctly interpreted IE Bulletin 79-08.
The actions taken cemonstrate the licensee's uncerstanding of the concerns arising f rom the THI-2 accident in reviewing their implementation on HNP 1 & 2 opera-tions, and provice added assurance for the protection of the public health and safety during the cperation of HNP 1 & 2.
1701 232
12 References 1.
IE Bu?letin 79-05, dated April 1, 1979, 2.
IE Bulletin 79-05A, dated April 5,1979.
3.
IE Bulletin 79-08, dated April 14, 1979.
4.
GPC letter, W. Widner to J. O'Reilly, dated April 25, 1979.
5.
GPC letter. W. Wicnar to J. O'Reilly, dated May 9, 1979.
6.
NRC staff letter, T. Ippolito to C. Whitmer, dated July 20, 1979.
7.
GPC letter, R. Kelly to T. Ippolito, dated August 10, 1979.
1701 233