ML19269D410
| ML19269D410 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/25/1979 |
| From: | Gallagher J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 7906020241 | |
| Download: ML19269D410 (16) | |
Text
. - }y s N
PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET TER$
P.O. BOX 8699 PHILADELPHIA. PA.19101 JOSEPH W. G ALLAGH ER ELECTRIC PRODUCT ON DEPARTMENT 12151 841-5o03 April 25, 1979 Docket Nos.
50-277 50-278 Re:
IE Bulletin 79-03 Mr. Boyce H.
Grier Director, Region I Office of Inspection and Enforcement U.
S.
Nuclear Regulatory Conmission 631 Park Avenue King of Prussia, PA 19406
Dear Str. Grier:
This correspondence is in response to NRC IE Bulletin No. 79-08.
The actions requested by the Bulletin and our responses are listed sequentially below.
1.
Review the description of circumstances described in Enclosure I of IE Bulletin 79-05 and the prelininary chronology of the TMI-2, 3/28/79 accident included in to IE Bulletin 79-05A.
a.
This review should be directed toward understanding:
(1) the extrene seriousness and consequences of the sinultaneous blocking of both trains of a safety system of the Three ::ile 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 automat'e action of engineered safety features unless continued operation of engineered safety features will result in unsafe plant conditions (see Section 5a of this Sulletin); and (2) not make 2260 332 00602 099 q
Mr. Boyce H.
Grier Page 2 April 25, 1979 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
The actions requested by item 1 of IE Bulletin 79-0S were covered by a program designed to inform the operational personnel of the known sequence of events pertaining to the TMI incident and the consequences thereof.
It also provided instruction in several areas of related operational concern.
The program was conducted in two phases, and i nclu d e c: all available licensed operators and menbers of the plant nanagement and supervisory staff having operational responsibilities.
Phase I consisted of the resident inspector holding direct discussions with licensed and'uulicensed operations and staff personnel and trainees on April 10-13, 1979, during day, afternoon and midnight shifts, with respect to details surrounding the events at TMI-2.
Licensed personnel from the corporate office were also in attendance.
The following topical areas were discussed:
(1)
An update of TMI-2 status (2)
A di=cussion of the six specific contributing factors to the incident as described on Pages 1 and ? of IE Bulletin 79-05A (3)
The seriousness a n.1 consequences of the sinultaneous blocking of both auxiliary feedwater trains (4)
The need for prompt reporting of serious events to the NRC and discussion regarding the licensee's reporting procedures (5)
The necessity to avoid prenature resetting of Engineered Safety Feature Systens, including core cooling systems and containment isolation systems.
These discussions also encoupassed resetting fron spurious signals.
It P C I and RCIC initiation was used as an example 2260 233
Mr. Boyce d.
Crier Page 3 April 25, 1979 (6)
The need to avoid premature tripping of Enginee red
'afety Feature Systems during any transients requiring flow The t.tation has maintained documentation regarding e t t e r.d a n c e at NRC discussions.
The resident inspector held diacussionc witn 77 personnel including 36 licensed oparators and senior operstors covering the above areas.
Staff, supervtsory, and nonlicensed individuals were also in attenden.e.
Addittonally, the station will provide the treining 9essions to those individuals who were not present during the i n s :) c e t i o n, upon their return to duty and to retata records cegarding this training.
Phase two consisted of formal instruction including 3roup dtscuncion directed by a nember of the plant supervisory staf: in accordance with a written format for all available licensed and operating supervisory personnel on a shift by shift basis during the week of April 9-13, 1979.
Dacnnentation of attendance has been retained.
Individuals who did not receive this training have been identified and
,ill be provided with sinilar training prior to May 1,
1979 or in the case of long term absences, before their resunption of licensed activities.
The training provided was neaningful and directed toward understanding:
(1) the seriousness and consequences of the simultaneous blocking of both trains of a safety system at the Three Mile Island Unit 2 (TMI-2) plant and other actions taken during the early phases of the accident; (2) the apparent operational errors which led to eventual core danage; and, (3) the necessity to systematically analyze plant conditions and parameters and take appropriate corrective action.
Specific areas addressed included:
(1)
Providing operators an awareness of the details of the Three Mile Island incident to the extent of information available at the time of this inspection. Additionally Bulletins 79-05 and 79-05A were made available in the control roon for inforuational purposes.
(2)
Reinstruction on specific measures which provide assurance that engineered safety features are available when required.
(~3)
Instructions on specific and detailed measures to assure that automatic actuations of energency safety features are not over-ridden.
2260 534
Mr. Boyce H.
Grier Page 4 April 25, 1979 (4)
Review of plant automatic actions initiated by reset of engineered safety features that could affect the control of radioactive liquids and gases.
(5)
Detailed on shift review of Procedures OT-1 " Reactor Low Water Level," E-14, " Loss of Coolant Accident (LOCA)," and CP-8 " Primary Containment Isolation."
2.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to initiate ccntainment 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 initiation design and procedures have been reviewed by site technical personnel and independently by the General Electric Company.
The review identified each line which penetrates the primary containnent, how it is isolated (automatically or uanually),
what signals initiate autonatic isolation, and whether it is needed in post accident conditions for plant safety or f or core cooling.
As a result of this review, three 3/4 inch lines in each plant were identified which are used during the performance of an Integrated Leak Rate Test (ILRT) of the primary containment.
These lines are triple valve isolated with the valves closed by the ILRT procedure.
In order to strengthen adninistrative control of these lines, one valve in each line has been blocked closed using a Shift Supervision Safety Block Pernit pending implinentation of pernanent positive controls.
Additionally, four small conductivity sample lines, one from the outlet of each RUR heat exchanger, which would carry post accident reactor water from the containnent to a sanple station in the reactor building and thence to the radwaste building were identified during the review.
The valves to isolate the line have been clearly identified with red paint.
The isolation procedure has been revised to ensure manual isolation during a Group I isolation which coincides with a high drywell pressure condition.
Engineering has been requested to review the design to determine if autonatic isolation of these lines is necessary.
2260
'35
Mr. Boyce H.
Grier Page 5 April 25, 1979 3.
Describe the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat renoval systens (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.
Response
Several automatic actions occur to ensure water delivery at high pressure to the reactor vessel whenever the main feedwater systen is not operable.
Loss of all feedwater delivery at full power would result in rapid le ve l 'd rop in the reactor vessel, resulting in a reactor scran at zero inches level (about 23 inches below normal operating level.)
As inventory continued to decrease, the HPCI and RCIC systens would receive an automatic start signal when reactor vessel level reached about ninus 48 inches.
Both of these high pressure systens would automatically start with no nanual operations required, and automatically inject water from the condensate storage tank to the reoctor systen.
Flow rates would be nominally 5000 GPM for HPCI and 600 GPM for RCIC.
Even if these systems had been pumping in its test loop, the initiation logic takes precedence over the operator, and all valves and controllers align thenselves for automatic delivery of water to the reactor systen.
The only exception to this occurs during a logic systen functional test, which takes approximately 1/2 hr and is performed every six nonths as required by plant Technical Specifications surveillance requireuents.
'4 h e n reactor water level has been re-established to nominally plus 45 inches, both HPCI and RCIC turbines would be autonatically shutdown.
The RCIC turbine would remain shutdown until the turbine trip throttle valve was reset from the nain coatrol room.
The HPCI turbine would again auto-start and inject water into the reactor vessel as soon as the lo level trip switch makes again at a vessel water level of ninus 48 inches.
Thus, high pressure delivery of water to the reactor vessel is assured with no operator actions recuired.
Even if the HPCI systen were manually shutdown, per the normal shutdown procedure, it would autonatically re-start and reflood the reactor vessel when level reached noninally ninus 48 inches which is about 11 feet over the top of the Core.
2260 ;36
Mr. Boyce H.
Grier Page 6 April 25, 1979 In the event that extended delivery of high pressure water to the reactor and venting of steam from the reactor via manual or autcmatic operation of the safety relief valves is required, torus cooling would be manually established via the RHR system in the torus cooling mode.
Even though the RHR system is in the torus cooling mode, it will automatically revert to the LPCI mode if required.
The core spray system would also start and deliver water to the ceactor vessel under the same conditions.
When the reactor has been depressurized, the RHR system can be placed in the long term shutdown cooling mode.
The low pressure ECCS functions are still operable, as required, even in the cold shutdown condition.
4.
Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systens.
Describe other redundant instrumentation which the operator m i,3ht have to give the same information regarding plant status.
Instruct operators to utilize other available inforuation to initiate safety systems.
.s
Response
Automatic initiation of safety systems based on reactor vessel water level is accomplished by the following instrument configurations:
a.
Reactor Protection System - control rod scram is accomplished by four analog loops arranged in a one-out-of-two twice logic such that the failure of any one switch or of any one set of s e n s i n,; lines will not defeat the safety action.
This schene utilizes two redundant GEMAC type head chambers and two redundant variable leg sensing lines.
b.
Primary Containment Isolation System - Group I isolation is accomplished by four analog loops arranged in a one-out-of-two twice logic such that the failure of any one switch or of any one set of sensin3 lines will not defeat the safety action.
This schene utilizes two redundant Yarway type head chanbers and two redundant variable leg sensing lines shared with the ECCS and ATUS functions.
c.
Emergency Core Cooling Systems - Initiation of ADS, HPCI, RCIC, LPCI, Core Spray and Diesel Generators is accomplished by four installed analog loops arranged in 2260 337
Mr. Boyce H.
Grier Page 7 April 25, 1979 a one-out-of-two twice logic for each of the initiation switches such that the failure of any one switch or of any one set of sensing lines will not defeat the safety action.
This scheme utilizes two redundant Yarway type head chambers and two redundant variable leg sensing lines shared with the PCIS and ATUS functions.
d.
Anticipated Transient Uithout Scram - trip of the reactor recirculation pumps on low reactor water level is accomplished by four separate mechanical switches arranged in a one-out-of-two logic for each punp such that failure of either switch or of either set of sensing lines will not defeat the safety action.
This schene utilizes two redundant Yarway type head chanbers and two redundant variable leg sensing lines shared with the PCIS and ECCS functions.
Manual initiation of the safety systems based on reactor vessel water level may also be accomplished by use of the following indications:
a.
Two redundant reactor water icvel indicators are installed on the reactor console.
These level instrunent loops are used for indication only.
This scheme employs the two redundant Yarway type head chambers and two redundant variable leg sensing lines utilized by the PCIS, ECCS and ATWS functions.
b.
As a back up to the level instrumentation in the control room, analog level loop indicators for the ECCS and PCIS functions are available on the local instrunent racks in the reactor buildings.
Other instrumentation including two separate sets of shutdown level instruments that the operator can make use of to deternine reactor plant status:
a)
Wide Range Shutdown Level - shown as an indicator on the ECCS panel - calibrated for cold conditions with indication from the nornal operating range to the vessel head vent.
b)
Narrow Range Shutdown Level - selectable on a recorder on the reactor console - calibrated for cold condations with indication from the top of the active fuel to 100 inches above the top of the active fuel.
c.
Reactor Vessel Pressure - there are five separate reactor pressure indications on the reactor console, 2260 338
Mr. Boyce H. Grier Page 8 April 25, 1979 three from the feedwater control system as indicators, one wide ra nge pressure displayed on a recorder and one narrow range pressure also displayed on a recorder.
d.
CRD Systen - flows and pressures are shown on indicators on the reactor console.
e.
RUCU System - flows are shown on indicators on a nearby control panel.
f.
Drywell Pressure
-a narrow range pressure indicator and recorder are on the ECCS control roon panels along with a high and low pressure annunciator.
Two redundant wide range pressure recorders are also mounted on the ECCS control room panels.
g.
Drywell Temperatures - a large number of thermocouples located throughout the drywell are available on the plant services panel on a push-button-select indicator.
There is one thernocouple which is used to continuously record drywell tenperature on the ECCS control roon panel.
h.
Drywell Sumps both integrators and recorders are used to provide sump punp out information for both the equipment and floor drain sumps.
Isolation valve position and punp operating status are shown on the sane panel.
Sump Hi-Hi level alarns are provided to indicate sump level abnormally high.
1.
Drywell Radiation - A continuous averaged sample from three elevations within the drywell is monitored by particulate, iodine and gaseous radiation detectors that alarm in the control room.
Actual values are continuously recorded at the local instrument rack in the reactor building.
J.
Torus Water Level - two narrow range devices (indicator and recorder) are provided on the ECCS control roon panel for normal operation.
Two redundant wide range level monitors are provided also on the same panel.
Alarms are provided to indicate when the level deviatcs from the narrow range predetermined limits.
k.
Reactor Building Radiatiai.
this parameter is monitored by several ares radiation monitors and by the ventilation exhaust radiation nonitors, all of which indicate and alarm on the control room radiation monitoring panels.
2260 539
Mr. Boyce H.
Grier Page 9 April 25, 1979 The requirement to instruct operators to utilize other available information to initiate safety systems is addressed under Iten 1 and Item 5.
5.
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 *(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 evaluating plant conditions.
Response
Piant operating procedures are being reviewed to ensure that they do not direct the operator to override automatic actions of engineered safety features unless continued operation of that feature will result in an unsafe plant condition.
The categories of operating procedures being reviewed are those which have the potential to direct the override of an engineered safety feature and include:
a.
plant emergency operating procedures b.
plant operational transient procedures c.
specific engineered safeguard systen operating procedures These procedural reviews will be conpleted within 2 weeks.
A letter was issued by the Station Superintendent, to all operating personnel re-enphasizing that all the redundant and conformatory instrumentation available be utilized when making operational decisions.
In addition, the plant administrative procedure which governs the conduct of shift operations is being revised to specifically address the concerns of ove rriding automa tic actions of safety features and making operating decisions based on observation of only a single paraneter during unusual plant conditions.
This procedure revision will be issued within one month.
2260 j40
Mr. Boyce H.
Grier Page 10 April 25, 1979 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 startup, and supervisory periodic (e.g.,
daily / shift checks,) surveillance to ensure that such valves are returned t 'o their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes.
Response
Safety related valve positions were verified during the week of April 9, 1979.
NRC inspectors accompanied by station personnel verified ESF systems valve / breaker / switch alignments by conducting the applicable system check-off lists for all of the accessible components in Units 2 and 3.
Accessible components considered were outside of the inerted primary containment, high radiation, high contamination or high airborne activity areas.
Check-off list completion verified the status of valves requiring positive position control (locking devices).
The following check-off lists were completed:
(1)
RHR (2)
CS (3)
ADS (4)
SBGTS (5)
SBLC (6)
HPCI (7)
RCIC (3)
Feedwater (9)
Condensate (10) Diesel Generator An independant reverification of safety related valve positions outside of primary containment has been completed by:
a.
performing valve check-off-lists on safety related system process valves in accordance with established procedures.
Those systems are defined as:
1.
High Pressure Coolant Injection 2.
Reactor Core Isolation Cooling 3.
Residual Heat Renoval 2260 341
Mr. Boyce H. Grier Page 11 April 25, 1979 4.
Core Spray 5.
Standby Liquia Control System 6.
Diesel Generator 7.
Standby Gas Treatment System 8.
Containment Atmospheric Dilution 9.
Emergency Service Water 10., High Pressure Service Water b.
performing instrument rack valve check-off-lists safety related instrunents in accordance with on established procedures.
The NRC inspector perforraed a comparison of the valve /
breaker /suitch aliganent procedures and check-off li'sts for the following EFf systens against current piping and instrument diagrams (P&ID's) and single-line diagrans to verify the adequacy of alignment procedures.
The following systens were reviewed:
(1)
RIIR/LPCI (2)
HPCI (3)
RCIC (4)
SBLC (5)
Core Spray (6)
SBGTS (7)
Electrical Breaker Alignnent for off-site power (8)
ADS (9)
PCIS (10) Diesel Generators (11) Secondary Containnent In addition, safety related system process valve positioning requirenents will be independently reviewed in depth by comparison between existing valve check-off-lists and the plant Piping and Ins t rumenta tion Diagrams.
The check-off-lists are a part of the appropriate systen procedures, surveillance tests and local leak rate tests.
This will be conpleted within three months.
Initial valve positions are presently governed by the appropriate procedural check-off-lists.
The controls which assure that va l ve s remain properly positioned are fundamen-tally three-fold:
a.
a locked valve log and associated administrative procedure b.
an adninistrative p roc e du re which sets requirenents for 2260 342
Mr. Boyce H.
Grier Page 12 April 25, 1979
\\,
post-work testing c.
a requirement in surveillance or routine testing procedures for a system or device to be returned to normal condition.
Review of these control mechanisms will be conpleted within three months.
A review will be conducted to determine that existing positive controls on safety related p roc e s s valves are adequate to assure these valves remain positioned to enable proper operation of the Safety Related Systens.
" Positive Controls" are interpreted to mean:
1.
Valves which are naintained in a known condition by means of locks which are under administrative controls.
2.
Valves which have position indications in the Control Room.
3.
Valves which are controlled by an interlock systen.
4 Valves which have position annunciation in the control roon.
The review will be completed within 3 months.
Movement of valves f rom an initial given condition occurs through the mechanism of systen operating procedures, and surveillance and routine testing procedures.
The procedures on safety-related systems are being reviewed to verify that a nechanisn exists to ensure that such valves are placed in their correct position following necessary nanipulations and are maintained in their proper positions.
7.
Review your operating nodes and procedures for all systens designed to t ra ns f e r potentially radioactive gases and liquids out of the p rima ry 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 ins t rumenta tion.
List all such systens and indicate:
a.
9hether interlocks exist to prevent transfer when high radiation indication exists, and 2260 343
Mr. Boyce H.
Grier Page 13 April 25, 1979 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
We are reviewing the design of-systems that could transfer potentially radioactive gases and liquids out of the primary containment and will verify that our current system designs and procedures will prevent the inadvertent release of radioactive liquids and gases.
The following are.the systems that have the ability to transfer potentially radioactive gases or liquids outside the primary containment:
Gas Systems Containment Ventilation and Purging System Drywell Oxygen Analyzer System Standby Gas Treatment System Containnent Atmospheric Dilution System CAD 0xygen and Hydrogen Analyzer Systems Main Steam Lines and Drains Reactor Core Isolation Cooling Steam Line High Pressure Coolant Injection Steam Line Liquid Systens Residual Heat Removal System Core Spray System High Pressure Coolant Injection Systen Reactor Core Isolation Cooling System Reactor Water Cleanup System Drywell Sump Systems Torus Water Cleanup System Primary Coolant Sampling Systens.
a.
No interlocks currently exist to prevent transfer of potentially radioactive gases or liquids when high radiation levels exist in the prinary containment.
However, the applicable procedures do require measurement of activity levels within containment prior to venting of containnent or resetting of isolation signals.
2260 344
Mr. Boyce H.
Grier Page 14 April 25, 1979 b.
All such systems are isolated by the applicable containment !cilation signals with the exception of the following:
1.
The containment Atmospheric Dilution System Oxygen and Ilydrogen Analyzer Systen which is required for post LOCA gas analysis.
2.
The Residual Heat Removal System sample valves.
An Engineering review request has been initiated to review the design to deternine if automatic isolation of these lines is necessary c.
The functional capability of the Primary Containment Isolation System is verified by surveillance testing every 6 months in accordance with the Technical Specifications.
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 froa service, b.
Verification of the operability of all safety-related systens when they are returned to service following maintenance or testing.
c.
Explicit notification of involved reactor operational personnel wheneve r a safety-related system is removed f rom and returned to service.
Response
The plant administrative procedure which governs corrective maintenance already addresses item Sb and has been revised to address item 8a.
Item ac for corrective naintenance is already acconnodated through the PECO Permit and Blocking system.
During the week of April 9, 1979, the NRC inspector reviewed the current, approved surveillance tests for the Engineered Safety Features (ESP) systems to verify that when the surveillance test is completed, the applicable systen will have been returned to an Operable condition.
Each applicable surveillance test was reviewed to insure that 2260 j45 1
i
Mr. Boyce H.
Grier Page 15 April 25, 1979 procedural steps were included that returned the system to I
an automatic initiation line-up.
2 The plant administrative procedure which governs f
surveillance testing will be revised to better address itens 8a and 8b.
Item 8e is already satisfactorily addressed.
Revision of administrative procedures will be completed r
within one month.
5 Safety-related surveillance and routine test procedures will be reviewed and revised as required.
A procedure sampling indicates that the items of section 8 have been already addressed.
The entire procedure review will be completed within three nonths.
9.
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time i
the reactor is not in a cont rolled o r expected condition of operation.
Further, at that time an open continuous connunication channel shall be established and maintained with NRC.
l
Response
I A letter has been..ssued to shift supervision and station 4
management personnel advising that prompt notification to
?
the NRC is required within one hour of the time that the reactor is not in a controlled or expected condition of operation and that a continuous conmunication channel nust j
be initiated at that time.
The letter lists the NRC i
emergency telephone nunbers.
A new administrative procedure will be issued to describe NRC prompt reporting to assure notification within one hour of the tine the reactor is not in a controlled or expected i
condition of operation.
It uill also identify that an open continuous communication channel be immediately established and na intained with the NRC.
This procedure will be written, approved, and inplemented by May 31, 1979.
10.
Revieu operating nodes 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.
Response
E' 2260 346
=
5 i
- f r. Boyce H.
Grier Page 16 April 25, 1979 Plant procedures related to LOCA's and CAD system operation have been reviewed.
These procedures adequately cover the control of combustible gases within containment.
The existing procedures are being revised to include a discussion of venting these gases from the RPV to the containment.
Multiple flowpaths currently exist to perform this operation.
11.
Propose changes, as required, to those technical specifications which must be modified as a result of your inplementing the items above.
Response
Correspondence on this item will be within the 30 days as delineated in the Bulletin.
Should you have any questions or require additional information, please do not hesitate to contact us.
Very truly yours,
. (,J.
_0l l
_1 2260 347