ML19296D523
| ML19296D523 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/14/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | Bauer E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| References | |
| IEB-79-08, IEB-79-8, NUDOCS 8003050019 | |
| Download: ML19296D523 (24) | |
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NUCLEAR REGULATORY COMMISSION 3
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February 14, 1980 Docket Nos. 50-277 and 50-278 Mr. Edward G. Bauer, Jr.
Vice President and General Counsel Philadelphia Electric Company 2301 Market Street Philadelphia, Pennsylvania 19101
Dear Mr. Bauer:
EUBJECT: NRC STAFF EVALUATION OF PHILADELPHIA ELECTRIC COMPANY RESPONSES TO IE BULLETIN 79-08 FOR PEACH BOTTOM ATOMIC POWER STATION UNITS 2 AND 3 We have completed our review of the information that you provided in your letters dated April 25, 1979 and May 14, 1979 in response to IE Bulletin 79-08 for the Peach Bottom Atomic Station, Units Nos. 2 and 3.
We have also completed our review of the supplemental infomation that you provided in your letter of August 8,1979.
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 of our evaluation is enclosed.
As you know, NRC staff review of the Three Mile Island, Unit 2 (TMI-2) accident is continuing and other corrective actions may be required at a later date. For example; the Bulletins and Orders Task Force is conduct-ing a generic review of operating boiling water reactor plants.
Specific requirements for your facility that result from this and other TMI-2 investigations will be addressed to you in separate correspondence.
Sincerely, q
f G
Thomas
. Ippo ito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Enclosure:
NRC Staff Evaluation
cc w/ enclosure:
See next page 8008080ofg
Mr. Edward G. Bauer, Jr.
Philadelphia Electric Company
-2 February 14, 1980 cc:
Eugene J. Bradley Philadelphia Electric Company Assistant General Counsel 2301 Market Street Philadelphia, Pennsylvania 19101 Troy B. Conner, Jr.
1747 Pennsylvania Avenue, N. W.
Washington, D. C.
20006 Raymond L. Hovis, Esquire 35 South Duke Street York, Pennsylvania 17401 Warren K. Rich, Esquire Assistant Attorney General Department of Natural Resources Annapolis, Maryland 21401 Government Publications Section State Library of Pennsylvania Education Building Commonwealth and Walnut Streets Harrisburg, Pennsylvania 17126 M. J. Cooney, Superintendent Generation Division - Nuclear Philadelphia Electric Company 2301 Marke' 'treet Philadelphia, Pennsylvania 19101 Edward Greenman U. S. Nuclear Regulatory Commission Office of Inspection and Enforcement Peach Bottom Atomic Power Station P. O. Box 399 Delta, Pennsylvania 17314 Philadelphia Electric Company ATTN:
Mr. W. T. Ullrich Peach Bottom Atomic Power Station Delta, Pennsylvania 17314
EVALUATION OF LICENSEE'S RESPONSES TO IE BULLETIN 79-08 PHILADELPHIA ELECTRIC COMPANY PEACH BOTTOM ATOMIC STATION, UNITS 1 AND 2 00CKET NOS. 50-277 AND 50-278
Introduction By letter dated April 14, 1979, we transmitted Office of Inspection and Enforce-ment (IE)Bulletin 79-08 to Philadalphia Electric Company (PECo 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 occurred at Three Mile Island, Unit 2 (TMI-2) on March 28, 1979.
By letter dated April 25, 1979, PECo provided responses to Action Items 1 through 10 of IE Bulletin 79-08 for the Peach Bottom Atomic Station, Units 2 and 3 (Peach Bottom 2 and 3).
PECo supplemented this response by a letter dated May 14, 1979 to provide the response to Action Item 11 of IE Bulletin 79-08.
The NRC staff review of the PEco responses led to the issuance of requests for additional information regarding the 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 8,1979, PECo responded to the staff's requests for additional information.
The PECo responses to IE Bulletin 79-08 provided the basis for our evaluation.
presented below.
In addition, the actions taken by the licensee in response to the bulletin and subsequent NRC requests were verified by onsite inspections by IE inspectors.
Evaluation Each of the 11 action items reauested 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.
1.
Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 March 28,1979 accident included in Enclosure 1 to IE Bulletin 79-05A.
a.
This review should be directed toward understanding:
(1) the extreme seriousness and consequences of the simultaneous blocking of both trains of a safety system at the Three Mile Island Unit 2 plant and other actions taken during the early phases of
. the accident; (2) the apparent operational errors which led to the eventual core damage; 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 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 operaticnal responsibilitips shall 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 25, 1979 described the program developed by PECo to inform the operational personnel of the known sequence of events pertaining to the TMI-2 incident and the consequences thereof.
It also
~
provided instruction in several areas of related operational concern.
The program was conducted in two phases, and included all available licensed operators and merrbers of the plant management and supervisory staff having operational responsibilities.
Phase one consisted of the resident inspector holding direct discussions with licensed and unlicensed operations and staff personnel and trainees on April 10-13, 1979, during the 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.
Phase two consisted of formal instruction including a group discussion directed by a member of the plant supervisory staff 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.
Documentation of attendance
. at both phases of these review and training sessions was maintained.
In its response, the licensee provided a detailed outline of the topics covered in each phase of the retraining and review program.
We conclude that the licensee's scope of review, instructions to operating personnel and documented participation satisfy the intent of IE Bulletin 79-08, Item 1.
2.
Review the containment isolation initiation design and procedures, and prepare and implement all changes neces:ary 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 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 initiation of 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.
In its response of April 25, 1979 the licensee stated that 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 containment, how it is isolated (automatically or manually), what signals initiate automatic isolation, and whether it is needed in post-accident conditions for plant safety or for core cooling.
As a result of this review, three 3/4-inch lines in each unit were identified whicn 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 administrative control of these lines, one valve in each line was blocked closed using a
. shift supervision safety block permit pending implementation of permanent positive controls.
Additionally, four small conductivity sample lines, one frem the outlet of each residual heat removal (RHR) system heat exchanger, which would carry post-accident reactor water from the containment to a sample 3tation in the reactor building and thence to the radwaste building, were identified during the review.
The valves to isolate these lines were clearly identified with red paint.
The isolation procedure was revised to ensure manual isolation curing a Group I isolation which coincides with a high drywell pressure condi-tion.
Engineering was requested to review the' design to determine if automatic isolation of these lines is necessary.
The licensee's letter of August 8, 1979 amplified on the above response, and included a tabulation of each fluid system and instrument line which penetrates primary containment.
Tne Peach Bottom 2 and 3 design, as other BWR designe, does not utilize
" initiation of safety injection" as a signal to initiate isolation of the primary containment.
Automatic isolation valves receive diverse actuation signals from the primary containment and reactor vessel isolation control system as described in Section 7.3 of the Peacn Bottom 2 and 3 Final Safety Analysis Report.
These isolation signals were listed and described in the aoove tabulation.
Automatic isolation is initiated for all. fluid lines which penetrate the primary containment whose isolation does not degrade needed safety features or cooling capability.
There are only four systems in this category.
The licensee discussed in detail the basis for not having automatic isolation of tnese lines.
Subseauently, the licensee reported that their engineering review has indicated tnat the four conductivity sample lines from the RHR system heat exchangers snould be provided with an automatic isolation feature.
A single solenoid valve controlled by the primary containment isolation system is planned for each sample line.
These valves will be seismically and environmentally qualified for this service.
Until these valves can be procured, installed and tested, the isolation procedure discussed above will remain in effect.
The goal is to have this accomplished on both units by mid-1980.
With respect to the three 3/4-inch ILRT lines discussed above, PECo provided additional information in its August 8, 1979 submittal on the actions taken to provide permanent positive control of these valves.
The licensee stated that provisions are being added to lock these valves closed.
A special block (supervision safety block permit) has been applied to the three ILRT lines.
Since the only time that these test valves would be used is during a refueling outage for the conduct of the ILRT, a procedural sign off step is being added to the controlling ILRT procedure to require application of the safety block permit and locking the valve closed for completion of the procedure.
The ILRT procedure must be complete prior to reactor start-up.
These valves will also be on the locked valve list.
We concluce that the licensee's review of containment isolation initiation design and procedure satisfies the intent of IE Bulletin 79-08, Item L.
3.
Describe 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 r..anual 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 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.
In its response of April 25, 1979, PECo stated that several automatic actions occur to ensure water delivery at high pressure to the reactor vessel whenever the main feedwater system is not operable.
Loss of all feedwater delivery at full power would result in rapid level drop in the reactor vessel, resulting in a reactor scram at the zero-inch level (about 23-inches below normal operating level.) As inventory continued to decrease, the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems would receive an automatic start signal when reactor vessel level reached about minus 48 inches.
Both of these high pressure systems would automatically start with no manual operations required, and automatically inject water from the condensate storage tank to the reactor system.
Flow rates would be nominally 5000 gallons / minute for the HPCI system and 600 gallons / minute for the RCIC system.
Even if these systems had been pumping in their test loops, the initiation logic takes precedence over the operator, and all valves a'nd controllers align themselves for automatic delivery of water to the reactor system.
The only exception to this occurs during a logic system functional test, which takes approximately 1/2 hour and is performed every six months as required by the plant Technical Specification surveillance requirements.
When reactor water level has been re-established to nominally plus 45 inches, both the HPCI and RCIC system turbines would be automatically shut down.
The RCIC system turbine would remain shut down until the turbine trip throttle valve was reset from the main control room.
The HPCI system turbine would again automatically start and inject water into the reactor vessel as soon as the low level trip switch makes again at a vessel water level of minus 48 inches.
Thus, high pressure delivery of water to the reactor vessel is assured with no operator actions requi^ d.
Even if the HPCI system were manually shut down per the normal shutdown procedure, it would automatically restart and reflood the reactor vessel when level reached nominally minus 48 inches which is about 11 feet over the top of the core.
In the event that extended delivery of high pressure wi.ar to the reactor and unting of steam from the reactor via manual or automatic 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 low pressure coolant injection (LPCI) mode if required.
The core spray system would also start and deliver water to the reactor vessel under the same conditions.
When the reactor has been depressurized, the RiiR system can be placed in the long term shutdown cooling mode.
The low pressure emergency core cooling system (ECCS) functions are still operable, as required, even in the cold shutdown condition.
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.
Describe all uses and types of vessel level indication for both automatic and manual initiation of safety systems.
Describe other redundant instru-mentation which the operator might have to give the same information 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) operators were instructed to utilize other available information to initiate safety systems.
In its submittal of April 25, 1979, PECo described the criteria covering automatic and manual initiation of safety systems based on reactor vessel water level and the other instrumentation available to the operators to determine plant status.
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 sensing lines will not defeat the safety action. This scheme 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 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 ECCS and anticipated transient without scram (ATWS) functions.
c.
Emergency Core Cooling Systems - Initiation of automatic depressurization system (ADS), HPCI, RCIC, LPIC, and core spray systems and the diesel-generators is accomplished by four installed analog loops arranged in 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 primary containment isolation syMem (PCIS) and ATWS functions.
d.
Anticipati.: Transient Without Scram - Trip to 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 pump such that failure of either switch or of either 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 ECCS functions.
Manual initiation of the safety systems based on reactor vessel water level may also be accomplished by use of the following incications:
.g-a.
Two redundant reactor water level indicators are installed on the reactor console.
These level instrument 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 backup to the level instrumentation in the control room, analog level loop indicators for the ECCS and PCIS functions are available on the local instrument racks in the reactor buildings.
Other instrumentation including two separate sets of shutdown level instruments that the operator can make use of to determine reactor plant status are as follows:
a.
Wide Range Shutdown Level - Shown as an indicator on the ECCS panel -
calibrated for cold conditions with indication from the normal operating range to the vessel head vent.
b.
Narrow Range Shutdown Level - Selectable on a recorder on the reactor console - calibrated for cold conditions 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 indica-tions on the reactor console, three from the feedwater control system as indicators, one wide range pressure displayed on a recorder and one narrow range pressure also displayed on a recorder.
c.
Control Rod Drive System - Flows and pressures are shown on indicators on the reactor console.
e.
Reactor Water Cleanup System - Flows are shown on indicators on a nearoy control panel.
. f.
Drywell System - A narrow range pressure indicator and recorder are on the ECCS control room 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 or thermocouples located throughout the drywell are available on the plant services panel on a pushbutton-select indicator.
There is one thermocouple which is used to continuously record drywell temperature on the ECCS control room panel.
h.
Drywell Sumps - Both integrators and reco'rders are used to provide sump pump out information for both the equipment and floor drain sumps.
Isolation valve position and pump operating status are shown on the same panel.
Sump hi-hi level alarms are provided to indicate sump level abnormally high.
i.
Drywell Radiation - A continuous averaged sample from three elevations within the drywell is monitored by particulate, iodine and gaseous radia-tion detectors that alarm in the control room.
Actual values are con-tinuously 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 room 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 deviates from the narrow range predetermined limits.
k.
Reactor Building Radiation - This parameter is monitored by several area radiation monitors and by the ventilation exhaust radiation monitors, all of which indicate and alarm on the control room radiation monitoring panels.
Ine requirement to instruct operators to utilize other available information to initiate safety systems is addressed under Item 1 and Item 5, above and below, respectively.
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:
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 additionpl 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.
The licensee's response was evaluated tu 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) that the review included operating procedures and training instructions.
In its response of April 25, 1979, the licensee reported that plant 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.
Subsequently, the licensee reported that this review has been completed.
The categories of operating procedures reviewed are those which have the potential to direct the override of an engineered safety feature and included:
Plant emergency operating procedures, Plant operational transient procedures, and Specific engineered safeguard system operating procedures.
A letter was iss.ed by the station superintendent to all operating personnel re emphasizing all the redundant and conformatory instrumentation available be utilized when making operational decisions.
In addition, the plant adminis-trative procedure which governs the conduct of shift operations has been revised to specifically address the concerns of overriding automatic actions of safety features and making operating decisions based on observation of only a single parameter during unusual plant conditions.
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 mar.~r 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 response was evaluated to assure that (1) safety-related valve positioning requirements were reviewed for correctress, (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 opp stion as well as during surveillance testing and maintenance.
In its response of April 25, 1979, the licensee stated that safety-related valve positions were verified during the week of April 9, 1979.
IE inspectors, accompanied by station personnel, verified engineered safety feature system valve / breaker / switch alignments by conducting the applicable system cneck-off lists for all of the accessible components.
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 cevices).
Check-off lists were completed for the following systems:
RHR system, Core spray system,
- ADS, Standby gas treatment system, Standby liquid control system, HPCI system, RCIC system, Feedwater system, Condensate system, and Diesel generator.
An independent 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 esta'blished procecures.
Those systems are defined as the following:
HPCI system, RCIC system, RHR system, Core spray system, Standby liquid control system, Diesel generator, Standby gas treatment system, Containment atmospheric dilution system, Emergency service water system, and High pressure service water system.
b.
performing instrument rack valve check-off lists on safety-related instruments in accorcance with established procedures.
75e IE inspectcr performed a comparison of the valve / breaker / switch alignment prc eaures and check-off lists for the engineered safety feature systems
. against current piping and instrument diagrams and single-line diagrams to verify the adequacy of alignment procedures. The following systems were reviewed:
RHR/LPCI systems, HPCI system, RCIC system, Standby liquid control system, Core spray system, Standby gas treatment system, Electrical breaker alignment for off-site power,
- ADS, PCIS, Diesel generators, and Secondary containment system.
In addition, safety-related system process valve positioning requirements will be independently reviewed in depth by comparison between existing valve check-off lists and the plant piping and instrumentation diagrams.
The check-off lists are a part of the appropriate system procedures, surveillance tests and local leak rate tests. We were subsequently advised by the licensee that this review has been completed.
Initial valve positions are presently governed by the appropriate procedural cneck-off lists.
The controls which assure that valves remain properly positioned are fundamentally three-fold:
a.
A locked valve log and associated administrative procedure, b.
An administrative procedure whicn sets requirements for post-work testing, and 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 has been completed.
. A review was also conducted to determine that existing positive controls on safety-related process valves are adequate to assure these valves remain positioned to enable proper operation of the safety-related systems.
" Positive Controls" are interpreted to mean:
a.
Valves which are maintained in a known condition by means of locks which are under administrative controls, b.
Valves which have position indications in the control room, c.
Valves which are controlled by an interlock system, and d.
Valves which have position annunciation in the control room.
Movement of valves from an initial given condition occurs through the mechanism of system operating procedures, and surveillance and routine testing procedures.
The procedures on safety-related systems were reviewed to verify that a mechanism exists to ensure that such valves are placed in their correct position following necessary manipulations and are maintained in their proper positions.
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 designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.
In particular, ensure that such an occurrence would-not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
a.
Whether interlocks exist to prevent transfer wnen 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 W ressed 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 its April 25, 1979 response, the licensee discussed the s,rstems that would have the capability of transferring potentially radioactive liquids and gases out of containment and advised us that the sys' tem designs and procedures were being reviewed to verify that the inadvertent release of radioactive materials would not occur.
The licensee did report that the applicable procedures required measurement of activity levels within containment prior to venting of containment or resetting of isolation signals.
The licensee's letter of August 8, 1979 reported that the review for the inadvertent release paths has been completed.
No pathways.were found which could inadvertently transfer either radioactive liquid or gas out of primary containment.
Those engineered safety features systems which are not required to initiate in the event of an accident are isolated and interlocked to prevent inadvertent transfer of radioactive liquids and gases.
Those engineered safety features systems which are required following an accident are not isolated.
In its response of April 25, 1979, the licensee reported that all such systems are isolated by the applicable containment isolation signals with the exception of the follawing:
a.
The containment atmospheric dilution system oxygen and hydrogen analyzer system which is required for post-loss-of-coolant accident gas analysis, and b.
The RHR system sample valves.
(An engineering review request has been initiated to review the design to determine if automatic isolation of these lines is necessary.)
The licensee's supplemental response of August 8, 1979 amplified the above, stating that no means were found which could result in an inadvertent transfer of radioactive liquid or gas due to the resetting of engineered safety features instrumentation.
The engineered safety features system designs are such that an action is carried to completion.
Resetting of the isolation requires manual operator action under controlled conditions following instrumentation reset.
In its response of April 25, 1979 and the supplemental response of August 8,
~
1979, PECo reported that the functional capability of the PCISs is verified by Technical Specification mandated surveillance f.ests as follcws:
a.
All of the initiating devices are functionally tested every month, and b.
Each logic system is functionally tested every six months.
We conclude that the licensee's review of systems designed to transfer radio-active 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 Bulletin 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 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 operabili'y of redundant safety-related systems is verified prior to the removal of any saf.ty-related system from service.
Where operability 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 c? the system status te.he 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 oncoming shift are explicitly notified during shift turnover about the status of systems removed from or returned to service since their previous shift.
The licensee's response of April 25, 1979 repo'rted that the plant administra-tive procedure which governs corrective maintenance already addresses Item 8b and has been revised to address Item 8a.
Item 8c for corrective maintenance is already accommodated through the PECo permit and blocking system.
During the week of April 9, 1979, an IE inspector reviewed the current, approved surveillance tests for the engineered safety features systems to verify that when the surveillance test is completed, the applicable system will have been returned to an operable condition.
Each applicable surveillance test was reviewed to insure that procedural steps were included that returned the system to an automatic initiation lineup.
The licensee's letter of August 8, 1979 amplified on the above response to Item 8a, reporting that it had revised tne " Procedure for Corrective Maintenance" to require that shift supervision verify by test or inspection the operability of redundant safety-related (Q-listed) systems required by the Technical Specifications.
Tne applicable test procedures have been reviewed and revised as necessary to require that any testing that could adversely affect the operability or reliability of a safety-related system is not permitted until verification of any redundent safety-related system is either done by test or inspection.
The licensee's letter of August 8,1979 also amplified its response to Item 8c by describing in detail the material that is covered in shift turnovers, the inspections that are conducted by the oncoming shifts, the use of tags to indicate system and equipment status and the meeting held by the shift super-intendent with all shift personnel at the beginning of each shift.
We conclude that the licensee's review and modification of maintenance, test and administrative procedures to assure the availability of safety-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 with the NRC following such events.
The licensee's response of April 25, 1979 reported that a letter has been issued to shift supervision and station 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 communication channel must be initiated at that time.
The letter lists the NRC emergency telephone umbers.
In addition, a new administrative procedure has been issued to describe NRC prompt reporting to assure notification within one hour of the time to the reactor is not in a controlled or expected condition of operation.
It also specifies that an open continucus communication channel be immediately established and maintained with the NRC.
We conclude that the licensee's response satisfies the intent of IE Bulletin 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.
In its response of April 25, 1979, the licensee advised us that plant procedures related to loss-of-coolant accidents and containment atmospheric dilution 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 reactor pressure vessel to the containment.
Multiple flowpaths currently exist to perform this operation. We were subsequently advised thct this review had been completed.
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 Items 1 through 10 of IE Bulletin 79-08.
The licensec reported in its letter dated May 15, 1979 that its review has shown that no cnanges to the Technical Specifications are required or anticipated.
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 the licensee to date, we conclude that the licensee has correctly interpreted IE Bulletin 79-03.
The actions taken demonstrate the licensee's understanding of the concerns arising from the TMI-2 accident in reviewing their implementation on Peach Bottom 2 and 3 and provide added assurance for the protection of the public health and safety during operation of the Peach Bottom Atomic Stati r,.
a References 1.
IE Bulletin 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.
PECo letter, J. W. Gallagher to B. H. Grier, dated April 25, 1979.
5.
PECo letter, S. L. Daltroff to B. H. Grier, dated May 14, 1979.
6.
NRC staff letter, T. A. Ippolito to E. G. Bauer, dated July 20, 1979.
7.
PECo letter, S. L. Daltroff to B. H. Grier, dated August 8,1979.