ML18024B214
| ML18024B214 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/27/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Parris H TENNESSEE VALLEY AUTHORITY |
| References | |
| IEB-79-08, IEB-79-8, NUDOCS 8001080028 | |
| Download: ML18024B214 (24) | |
Text
Docket Nos. 50-259 5
296 DECEt~lHER 2 V 1979 tlr,. Pugh 6. Parris Haqager of Power Tehnessee Valley Authority 500A Chestnut Street, Tower II Chattanooga, Tennessee
'3?401 Dear tlr. Parris; Distribution
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SUBJECT:
NRC STAFF EVALUATION OF TENNESSEE VALLEY AUTHORITY RESPONSES TO IE BULLETIN 79-08 FOR BROIIJHS FERRY NUCLEAR PLANT Me have completed our review of the information that you provided fn your letters dated April 24 and July ll. 1979 fn response to IE Bulletin 79-08 for the Browns Ferry Nuclear-Plants.
(le have also comp'teted our review of the, supplemental fnfermatfon that, you provided in your letter of August 6, 1979 We have concluded that you have taken the appropriate actions to meet the requfretrents of each of the eleven action items identified.in IE Bulletin 79-08.
A copy of our evaluation fs enclosed.
As you know, NRC staff revfew of the Three tlile Is'fang, Unit 2 (THI-2) accident is continuing and other corrective actions may be required at a later date.
For example. the Bulletins and Orders Task Force is
'onducting a generic revfeii of operating boiling water. reactor plants.
Specific requfreiItents for your facility that result from thfs and other TthI-2 investigations will be addressed to you in separate correspondence.
Sincerely, V.P Thomas A. Ippolfto, Chief Operating Reactors Branch P3 Division of Operating Reactors
Enclosure:
Stai'f Evaluation cc w/enclosure; See next page
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NRC FORM 318 (9 76) NRCM 0240 AV.S. GOVERNMENT PRINTING OFFICE: 1979 289'369
0 S
hy 'lf
0 Hr. Hugh G. Parris
=Tennessee Valley Authority CC:
H.
S, Sanger, Jr,,
Esquire General Counsel Tennessee Valley Authority 400 Commerce Avenue E11833C Knoxville, Tennessee 37902 ter.
Ron Rogers Tennessee Valley Authority 400 Chestnut. Street, Tower II Chattanooga, Tennessee 37401 E.
G. 8easlev Tennessee Valley Authority 400 Commerce Avenue M 10C 131C Knoxville, Tennessee 37902 Robert F. Sullivan U.
S. Nuclear Regulatory Commission P.
0.
Box 1863
- Oecatur, Alabama 35602 Athens Public Library South and Forrest
- Athens, Alabama 35611
EVALUATION OF LICENSEE' RESPONSES TO IE BULLETIN 79-08 TENNESSEE YALLEY AUTHORITY BROMNS FERRY NUCLEAR PLANT, UNIT NOS.
1, 2
AHO 3 DOCKET NOS.
50-259, 50-260 and 50-296
Introduction By letter dated April 14,
- 1979, we transmitted Office of Inspection and Enforcement (IE) Bulletin 79-08'o the Tennessee Valley Authority (TVA or the licensee).
IE Bulletin 79-08 specified actions to be taken by the licensee to avoid occurrence of an event similar to that which occur red at Three Mile Island, Unit 2 (TMI-2) on March 28, 1979.
By letter dated April 24,
- 1979, TVA provided responses to Action Items 1 through 10 of IE Bulletin 79-08 for the Browns Ferry Nuclear Plant, Units Nos.
1, 2 and 3 (BF-l, BF-2 and BF-3).
By leiter dated July ll, 1979, TVA provided its response to Action Item 11.
The NRC staff review of the TVA responses led to the issuance of requests for additional information regarding TVA's responses to certain action items of IE Bulletin 79-08.
These requests were contained in a letter dated July 20, 1979.
By letter dated August 6,
- 1979, TVA responded to the staff's requests for additional information.
The TVA responses to IE Bulletin 79-08 provided the basis for our evaluation pr esented below.
In addition, the actions taken by the licensee in r sponse to the bulletin requirements and subsequent HRC requests were verified through onsite inspections by IE inspectors..
Evaluation Each of the ll action items requested by IE Bulletin 79-08 is repeated below followed by our criteria for evaluating the response, a summary of the licensee's response and our evaluation of the response.
Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of ihe THI-2 March 28, 1979 accident included in Enclosure 1 to IE Bulletin 79-05A.
This review should be directed icward understanding:
(1) 'he extreme seriousness and consequences of ihe simultaneous blocking of both trains of a sa eiy sys-em ai the Three
!>i'~e Is land Un".2 plant and oiher actions iaken during the early phases of ihe acciden';
(2) the appareni operaiional errors
b.
which led to the eventual core damage; and (3) the necessity to systematically analyze plant conditions and parameters and take appropriate cor rective action.
Operational personnel should be instructed to (1) not override automatic action of encineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a 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 operational responsibilities'hall participate in this review and such participation shall be documented in plant records.
The licensee's response was evaluated to determine that (1) the scope 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 of April 24, l979 discussed the review and instructions that were being presented to all of TVA's managers and super-visors with operational responsibilities and to all licensed operators.
This program was scheduled to be completed by May 27, l979.
The outline of the review covered all of the points raised in Action Item l above.
TVA committed to document the contents of and participation in this review.
We conclude that the licensee's scope of review, instructions to operating personnel and documented participation satisfies the intent of IE Bulletin 79-08, Item l.
2.
Review the containment isolation initiation design and procedures, and prepare and'mplement all chances 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.
The licensee's response was evaluate"
-o verify that containment isolation i nitiation design and procedures hac been reviewed to assure that (1) manual
0 0
or automatic initiation of containment isolation occurs on automatic initia-tion of safety injection and (2) all lines (inc1uding those designed to transfer radioactive gases or liquids) whose isolation does not degrade cooling capability or needed safety features were addressed.
The licensee's response of April 24, 1979 stated that it had reviewed the containment isolation design and procedures to ensure that proper isolation of the primary containment occurs upon initiation of emergency core cooling systems (ECCS).
All containment isolation valves except those needed for ECCS opera ion will automatically close on receipt oi a primary containment isola-tion signal (PCIS).
The PCIS is generated by high drywell pressure or by reactor low level.
The PCIS setpoints are chosen such that isolation will occur prior to or at the same time that ECCS initiates.
The logic design is diverse and redundant, ensuring high probability of completing its intended function.
There are several other isolation modes in addition to the main PCIS logic.
For example, main steam line isolation valves will also close as a result of high steam line radiation, high steam flow, or high steam line tunnel t,emperature.
The primary containment ventilation system isolates on reactor building high radiation.
The high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems have instrumentation to detect pipe breaks within their own flow paths, and to subsequently isolate the system.
In a supplemental response dated August 6,
- 1979, TVA confirmed that its review considered initiation of containment isolation of all lines penetrating containment (including those designed to transfer potentially radioactive gases and liquids out of containment) on receipt of a PCIS.
The PCIS setpoints are chosen so that isolation will occur before or at the same
'ime that ECCS initiates.
The licensee further stated thai no changes to design or procedures are necessary to mee he requirements of I= Bulletin 79-08.
Me conclude that the licensee's review of containmenz, isolation initiation design and procedures satisfies the intent of IE 8ulletin 79-08, Item 2.
3.
Oescribe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal systems (e. g.,
RCIC) that are used when 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 sense.
The licensee's response was reviewed to assure that (1) it described the automatic and manual actions necessary for ti'ie proper func ioning of the auxiliary heat removal systems when the main feedwater sys em is not operable and (2) the procedures for any necessary manual actions were described in summary form.
In its letter of April 24,
- 1979, TVA stated that on a loss of feedwater flow the following event, sequence is initated automatically; no operator action is required other than to verify automatic system operation as described below.
If automatic functions do not occur, procedures require that systems be manually initiated.
Core recirculation flow reduces to minimum, reactor scrams on low water level, main steam isolation valves close, RCIC and HPCI systems go into operation.
As a result of reactor isolation, pressure increases causing the relief valves (RV's) to lift and discharge to the pressure suppression
- pool, a 985,000-gallon passive neat sink.
Either system, RCIC or HPCI, is capable of maintaining adequate reactor water level as reactor pressure is controlled through the RV's.
If one or more RV's fail to
- close, the HPCI and RCIC systems will continue to maintain adequate reactor water level as primary system pressure decreases.
Continued depressurization will enable automatic operation of the low pressure core spray (LPCS) and low pressure coolant injection (LPCI) sys ems, either of which is capable of maintaining reactor water level.
In addition, other high and low pressure systems are available to maintain reactor water level and can be initiated manually.
These include the residual heat removal service water system, the standby liquid control
- system, and the control rod drive system.
All emergency operating instructions have been reviewed to ensure that specific instructions and requirements for manual initiation of each safety system is provided.
This manual initiation i's immediate following the failure of an automatic initiation.
We conclude that the licensee's procedural summary of automatic/manual actions necessary for the proper functioning of auxiliary heat removal systems used when the main feedwater system is inoperable satisfies the intent of IE Bulletin 79-08, Item 3.
- 4. 'escribe all us'es and types of vessel level indication for both automatic and manual initiation of safety systems.
Describe other redundant in'strumentation which the operator might have to give the same informa-tion regarding plant status.
Instruct operators to utilize other available 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 addressed 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 response of April 24, 1979 stated that the
", ange of reactor water level from 100 inches below the top of the active fuel up to the
-:op of the reactor vessel is co'vered by a combination of narrow and wide range instruments.
There are two basis types of level instrumentation:
(1) level transmitters which require excitation voltage to function and (2) level indicating switches which do not require power for switch operation.
The safety-related systems or functions served by reactor water level instrumentation are listed below.
Function or S stem Reactor protection system
- trips reactor Number of Instruments 4 Barton level indicating switches 4 Yarway level indicating switches Primary containment and reactor isolation Core and containment cooling initiation
- CSS, LPCI, and AOS 4 Yarway level indicating switches ADS 2 Yarway level indicating switches Containment spray 2 Yarway level indicating switches In all cases redundancy is provided to ensure the logic can perform its intended function.
This includes alternate and redundant power sources for initiation logic.
Separation of logic trains is maintained by separation of i nstrument sensing lines, instrument panels, cable trays, and instrument room panels.
The entire installation is seismically qualified.
Test~ of boiling water reactor water level instrumentation under simulated steam and water line breaks have been conducted showing satisfactory performance.
In addition to the above functions, there are three GEMAC level transmitters which provide input signals to the feedwater control system and trip the main turbine and feedpump turbines on high water level in the reactor.
There are three additional level indicators (two Yarway indicators and one GEHAC transmitter) which perform no control or protection functions;
- however, these level indicators are available to the operator.
There are a total of 22 instruments for measuring
Eight of these instruments have indicators and/or recorders in the main control room.
The diversi ty and rec undancy of these instruments assure adequate information is provided to automatically initia-e safety actions and provide the operator with assurance of the vessel water level at all times.
In its response of April 24,
- 1979, and reconfirmed in its response of August 6,
- 1979, TVA stated that a retraining program for all operators had been completed in which it, was emphasized that they should use other confirmatory sources of information to initiate safety systems.
The training program emphasizes the interpretation of all available information in order that the operator can diagnose the basic cause of any malfunction or abnormal occurrence.
The plant normal and emergency operating instructions list specific confirmatory indications and expected system parameter changes associated with equipment malfunctions or postulated accidents.
The prescribed operator response to the abnormal situations also lists the confirmatory indications to verify appropriate corrective action is being taken.
Me 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 8ulletin 79-08, Item 4.
5.
Review the actions 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 (e.g.,
vessel integrity).
b.
Operators 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
'improperly 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) ;hat the rev,ew included operating procedures and sr aining instructions.
In its response of April 24,
- 1979, TVA stated that the operating procedures and training instructions were being reviewed to ensure that operators are instructed not to over ride automatic operations of the engineered safety
- features, unless continued operation of the engineered safety system will result in unsafe plant conditions, or until the plant is clearly under control and the engineered safeguards are no longer required.
This review was to have been comp'leted by Hay 25, 1979.
In its response of April 24,
- 1979, and reconfirmed in its response of August 6,
- 1979, TVA reiterated its response in Action em 4 above and stated that all operating instructions and training instructions had been reviewed and revised as necessary to instruct operators not to rely on vessel level indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.
TVA also pointed out that in addition to the multiple level indications described in the response to Action i.tern 4 above, there are over a dozen other types of instrumentation in the boiling water reactor to provide the operator with indirect ind'tion of reactor vessel coolant inventory changes and could inform the operator of the need to take corrective actions.
Examples are listed below.
Orywell High Pressure Orywell High Radioactivity Levels Suppression Pool High Temperature Relief Valve Oischarge High Temperature High Feedwater Flow Rates High Hain Steam Flow High Containment and Equipment Area Temperatures High Oifferential Flow - Reactor Mater Cleanup System Abnormal Reactor Pressure High Suppression Pool Mater Level H gh Oryweli and Containment Sump Fill and Pumpou-Rate Valve Stem Leakoff High Temperature High Process Honitor Radiation Levels
Abnormal Reactor Recirculation Flow High Electrical Current (Amperes) to Pump Motors Me 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 ".equirements 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 relat'ed 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 ail operat'onal modes.
The licensee's response was evaluated to assure that (1) safety-related valve positioning requirements wer e 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.
In its initial response of April 24, 1979 and in i s supplemental response of August 6,
- 1979, TVA reported that it had completed a.eview of safety-related valve positioning requirements to ensure proper operation of engineered safety features.
TVA noted that plant administrative procecures require that (a) all critical safety system and component (CSSC) alignment is verified prior to unit startup, (b) changes in the alignment of any CSSC is recorded on a system status sheet and (c) shift personnel being relieved communicate information on any abnormal plant condition including temporary conditions.
TVA also reported that CSSC valve position requirements are listed and documented in the operating instruction valve checklist for each individual system.
The complete system valve checklist is done before sta tup following each refueling outage and placed in the system status files.
TVA's response stated that locked safety system va',ves were included on sys-em valve checklists.
During an HRC site inspection in ivay of 1979, several ECCS
10 valves listed on TVA's valve checklists as locked were found to be unlocked.
This was identified as an item of noncompliance in the NRC letter from R.
C.
Lewis to H.
G. Parris dated August 7, 1979.
As part of the corrective action to this item of noncompliance, TYA revised al.l ECCS valve checklists to include second-party verification of the valves.
This action was verified completed on November 2, l979.
The TYA response also reported that if the position of any locked valve is required, to be changed, an "abnormal status" form is filled out and placed in the status file for the appropriate system.
The form states the normal and abnormal condition and the reason for the abnormal status.
Should this change render the system or component inoperable, further action is di rected as outlined in response to Action Item 8 below.
Mhen the locked valve is returned to the normal position, the abnormal status sheet is removed from the file.
The file is reviewed for completeness and abnormal status by a weekly status review procedure.
Major flow path valves for each system are checked on a separate panel checklist for proper position weekly.
Acces ible safety-re'lated valves in the main flow path of CSSC equipment have been inspected to verify proper positioning.
This verification is performed by a weekly status review procedure.
Me 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.
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.
Mhether interlocks exist to prevent transfer when high radiation indication exists, and
b.
Whether such systems are isolated by the containment isolation s i gnal c.
The basis on which continued operability of the above features is assured.
The licensee's response was evaluated to determine that (i) 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.
At the Browns Ferry units, there are two potential sources by which radioactive water could be transferred from the reactor containment building to the auxi lliary bui lding - the drywel 1 drai n sump and the drywel 1 equipment drain sump.
The primary function of the drywell drain sump and drywell equipment drain sump is to collect and measure nuclear system leakage from both identified and unidentified sources.
Two pumps associated with each sump operate automatically to maintain the sump level within a desired range.
These pumps discharge to the radwaste system.
Two isolation valves in each discharge line are present.
These valves fail in the closed position on loss of power or control air.
in its letter of April 24,
- 1979, TVA stated that while there are no high radiation interlocks associated with the control logic on the drywell drain sump and drywell equipment drain sump isolation valves, these valves will isolate on receipt of a PCIS.
The PCES is initiated by either low reactor
~ater level or high drywell pressure.
The response also stated that once the PCIS is initated, manual operator aciion is required to permit reopening of ihese discharge valves.
The unit operator may manually secure the drain pumps if high radi ation is suspected.
The primary containment has area radiation monitors thai would alert the unit operator of high containment aciivity.
The radwaste building has similar monitors with annunciaiors local io ihe radwaste
12 area and also in the main control room.
Annunciators wi'il also alarm if either transfer pump operates for an excessive time (about ten minutes).
The radwaste factility is continuously manned and these operators can isolate high activity in-flow from their control station as well..
TVA's letter of August 6,
- 1979, which was in response to our request for additional information, stated that it intends to install detectors on these sump lines to interlock sump pump operation when high radiation is present in the sump line effluent.
In the meantime, plant operating instructions require
- that, before resetting a PCIS signal, an evaluation be made to ensure that inadvertent transfer of significant amounts of contaminated fluids will not occur.
In a telephone conversation with TVA on October 18,
- 1979, we requested additional information as to what the procedures specifically require.
The procedures require that before the operators clear a PCIS signal, they must personnally check that each isolation is manually closed or clos it if it is not closed.
Also, before the manual isolation valves are opened and the sump pumps restored to operation, the sump water must be analyzed for activity.
Based on the procedures which TVA has instituted and the inter locked radiation monitors which they plan to install, we conclude that there is reasonable assurance that water containing a significant level of radioactivity will not be inadvertently transferred out of containment.
Mith regard to the transfer of gases into and out of containment, the primary containment purge (PCP) system is used during reactor startup to nitrogen inert the containment, or during reactor shutdown to provide a breathable atmosphere.
During routine power operation, the small vents on this system are infrequently used to add makeup nitrogen or to adjust drywell pressure.
The ventilation valves are maintained closed unless the sys'em is actually in
13 The PCP system normally processes the exhaust streams through HEPA filters and charcoal filters into the reactor building ventilation system.
Alignment to standby gas treatment is also possible.
The exhaust gases from the reactor and turbine building are continuously monitored for radioactivity, and annunciators would alert the unit operator well before any Technical Specification release rates are approached.
In its responses of April 24, 1979 and August 6, 1979; TVA stated that the primary containment ventilation valves will automatically isolate (if open) on receipt of the PCIS signal (high drywell pressure or low vessel level).
Additionally, the valves will isolate and cannot be reopened if high radiation is detected in the reactor building ventilation ducts.
Limited PCIS override controls exist that provide operations personnel with a preplanned method of releasing gas from either the drywell or pressure suppression pool to the standby gas treatment system for long-term post-LOCA atmospheric control.
Me conclude that the licensee's review of systems designed to tr"nsfer 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 8ulletin 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 cf the operability of safety-.elated 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.
The licensee's response was evaluated to determine that operability of recundant safety-reia-ed systems is verified prior to tne removal of any safety-related svstem from service.
Where operabi',ity verification appeared
only to rely on previous surveillance testing within Technical Specification intervals, we asked that 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 checked to see that all involved reactor operational personnel in the oncomi ng shift are explicitly notified during shift turnover about'he status of systems removed from or returned to service since their previous shift.
In its response of April 24, 1979,'TVA stated that plant specific administrative procedures require:
(a) verification of the operability of redundant s'afety-related equipment before such equipment is removed from service (equipment operability requirements are based on plant Technical Specifications),
(b) that system operability is proven before a system is returned to service and (c) approval by the shift supervisor or his representative prior to the performance of any activity on p'lant equipment or any activity that may affect plant equipment.
In addition, the shift supervisor or his. representative is notified when an activity authorized to be performed on plant equipment is completed or a change occurs in the scope of the activity.
In its supplemental response of August 6,
- 1979, TYA amplified on (a) above
~o point out that prior operability verification within the current Technical Specification surveillance interval is not relied on.
Plant specific administrative procedures, based on Technical Specifications, require that redundant safety system(s) be proved operable by performance of applicable surveillance instruction(s) before removing a safety system from service.
- he above position is reflected in plant maintenance and est procedures and exceeds the proposed requirement to perform a visual check of system status.
n its initial response of April 24,
- 1979, TVA -also stated that in addit.'on o
prewri tten ins'ruc ions, maintenance activities are con rolled through a
trouble-reporting sys
.em which identifies specific maintenance requirements on
15 each trouble report (TR).
Before maintenance is performed, the work require-ments of the TR are reviewed by the shift supervisor or his representative.
Approval to perform maintenance is indicated by signature o
the shift supervisor on each TR.
After completion of work, the shift supervisor or his representative is notified that the maintenance activity is complete.
Return to normal instructions and test requirements are specified in a referenced maintenance or surveillance instruction or on the initiating TR.
TVA's supplemental response of August 6, 1979 amplified the above to explain how reactor operational personnel from an oncoming shift are notified about the status of systems removed from or returned to service.
Plant procedures require that systems removed from or returned to service be documented in the operator's daily journal and that each oncoming operator read the journal back to the last shift worked or five calendar
- days, whichever is less.
Me conclude that the licensee's review and modification of maintenance, test and administrative procedures to assure the availability of saA "y-related 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.
The 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 the time the reactor is not in a controlled or expected condition of operation and (2) procedures required or were to be modified to require the establishment and maintenance of an open continuous communication channel wi:h the NRC following such events.
In its response oi April 24,
- 1979, TVA stated that the required report,)ng procedures would be incorporated in the plant operating procedures.
Me conclude that the licensee's response satisfies the intent of EE Sulletin 79-08, Item 9.
10.
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.
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.
During normal operation, radiolytically produced hydrogen and oxygen are vented from the primary coolant to the condensor and are subsequently recombined in the catalytic beds.
Mhen the reactor is shut down, minimal hydrogen is evolved.
In its response of April 24,
- 1979, TVA stated that it had further reviewed the operating modes and procedures to deal with significant amounts of hydrogen that might be generated during a transient or accident.
Features are available for safe handling and disposal of the hydrogen.
.To remove hydrogen from an isolated reactor vessel, the main steam relief valves may be remotely operated to vent the hydrogen to the suppression pool.
In addition, the reactor vessel has a head vent line with valves remotely operated from the control room.
However, the valves which vent to containment are not normally opened without first depressurizing the system.
The primary containment is normally inerted with nitrogen to maintain low oxygen concentration.
Redundant instrumentation is available to continuously monitor both the oxygen and hydrogen concentrations in the torus and drywell.
Nitrogen stored in tanks as part of the containment atmospheric dilution (CAD) system is available for control)ing hydrogen and oxygen concentrations.
Existing operating procedures adequate!y prescribe the operation of the CAD system to prevent buildup of detonatable gases in the primary containment.
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 be 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 Action Items 1 though 10 of IE Bulletin 79-08.
By its letter dated July ll, 1979, the licensee advised us that its review has shown that no changes to the Technical Specifications are required.
The licensee also advised us that in its continuing review, should modifications to the Technical Specifications be required, 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 Based on our review of the information provided by tne licensee to date, we conclude that the licensee has correctly interpreted IE Bulletin 79-08.
The ac ions taken demonstrate the licensee's understanding of the concerns arising from the THI-2 accident in reviewing their implementation on BF-1, BF-2 and BF-3 operations, and provide added assurance for the protection of the public health and safety during the operation of the Browns Ferry Nuclear Plant.
References IE Bulletin 79-0:,
dated April 1, l979..
~.
18 2.
IE Bul letin 79-05A, dated April 5, 1979.
3.
IE Bulletin 79-08, dated April 14, 1979.
4.
TVA letter, J.
E. Gilleland to J.
P. O'Reilly, dated April 24, 1979.
S.
TVA letter, L.
M. Mills to J.
P. O'Reilly, dated July 11, 1979.
6.
NRC staff letter, T.
A. Ippolito to H.
G. Parris dated July 20, 1979.
7.'VA letter, L.
M. Hills to T.
A. Ippolito dated August 6, l979.
8.
RII Inspection Report Nos.
50-259/79-14, 50-260/79-14 and 50-296/79-14.
9.
TVA's response to the item of noncompliance L.
M. Mills to J.
P.
O'Reilly dated August 29, 1979.
10.
RII Inspection Report Nos.
50-259/70-33, 50-260/79-33 and 50-296/79-33.