ML19309D518
| ML19309D518 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 04/04/1980 |
| From: | Andognini G BOSTON EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 80-54, NUDOCS 8004100414 | |
| Download: ML19309D518 (35) | |
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BCETON EDISON COMPANY a ca s m *6 orricas eco sovssTom steer 7 Bostow. MassAcMussTTs a2199
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s u P( 48N T E ND E N T NUCLEAR optsaf oms OEPARTMENT April 4, 1980 BECo. Ler. #80-54 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D. C.
20555 License No. DPR-35 Decket Number 50-293 Implementation Details of Lessons Learned Short Term Requirements
References:
A) NRC Letter from Mr. H. R. Denton to All Operating Nuclear Power Plants, dated October 30, 1979 Discussions of Lessons Learned Short Term Requirements B) BECe Letter Number 79-241 dated November 21, 1979, Response to Discussion of Lessons Learned Short Term Requirements C) BECo Letter Number 79-206 dated October 19, 1979, Follow-up to Reviews Regarding the Three Mile Island Unit 2 Accident D) BECo Letter Number 79-281 Response to Lessons Learned Short Term Requirements
Dear Sir:
Reference A) requested that all operating nuclear power plants provide a detailed description and justification for each of their responses which were not in com-plete agreement with your staff's requirements as clarified by that reference.
Boston Edison Ccmpany performed a preliminary review of those clarifications and responded in Reference B).
Additionally, Reference A) required Bosten Edison Company to submit to the NRC by January 1,1990, a detailed description and schedule fer the following five NUREG-0573 positions:
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CDOTON EDCON COMPANY Mr. Harold R. Denton, Director April 4, 1980 Page 2 (1)
Position 2.1.1 Emergency Power Availability to Provide Long Term Source of Air for Air-Operated Relief Valves (2) Position 2.1.2 Relief and Safety Valve Testing (3) Position 2.1.5.a -
Dedicated H2 Control Penetrations (4) Position 2.1.8.a -
Post-Accident Sampling (5) Position 2.2.2.b -
On-Site Technical Support Center Our responses to those positions were forwarded to you by Reference D).
Reference D) also stated that we would provide you with our methods of implementation of Category A items before the beginning of our next operating cycle.
At a meeting with the NRC's BWR TMI Task Force on March 18, at Pilgrim Station, BECo was requested to provide methods of implementation of Category A items in advance of our restart date to allow NRC ample time to review and approve the submittal prior to restart.
Based on our discussions and subsequent telephone conversations with the Task Force and Mr. Verrelli, we understand that our modifications and this submittal as described below conform to all of the NRC's Category "A". requirements.
In addition, we also understand that, to the extent that Category B requirements are described, they are also in full conformance to the NRC's requirements.
Very truly yours,
/
o 2.1.1 Emergency Power Availability to Provide Long Term Source of Air for Air-Operated Relief Valves We understand that our submittal in Reference D) BEco Letter #79-281 dated December 31, 1979 was adequate to meet the NRC's concerns.
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2.1.2 Performance Testing of PWR and BWR Relief and Safety Valves Boston Edison Company endorses the position of the BWR Owners' Group on Safety and Relief Valve Testing. General Electric Company is presently at work pre -
paring a total program plan; including the scope of testing, analysis to be performed, and a schedule for completion.
We understand that this is satisfactory to the NRC staff requirements.
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2.1.3.a Direct Indication of Power-Operated Relief Valves and Safety Valve Position for BWR's All spring safety and safety / relief valves at Pilgrim Station Unit #1 will be equipped with aesustic monitoring devices to insure that operations personnel have a positive indication of flow in the discharge line or tail pipe assembly.
The acoustic monitoring systen will consist of:
a) Acoustic transducers-One on each S/R Valve discharge line and spring safety tail pipe.
b) Pre-amplifier modules - One for each transducer to be located in close proximitry within the primary containment.
c) Amplifier modules - One for each sensing unit and located on r.ne Post Accident Monitoring (PAM) Panel to be located in the control room.
(C171) d) An LED Display - Located on the PAM Panel to indicate percent of open for each valve.
e) Annunciator - One common to alarm on valve open signal.
f) Power - System power will be 120 VAC from a Class 1E emergency power source.
At present, environmental and seismic qualification conforming to IEEE-323 1974 is in process for system components.
A secondary system to provide back-up indication is provided by temperature monitoring with thermocouples installed on each discharge or tail pipe. This system utilizes a temperature recorder and a common annunciator, both in the control room, and provides signal input to the plant computer. Power for this system is provided from a Class 1E source independent of the source for the primary system (i.e., accoustic monitoring).
Plant procedures detailing operation of both systems will be implemented.
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l 2.1.3.b Instrumentation for Core Cooling Studies by GE to date, indicate that improved RV level instrumentation is recom-mended. At this time the final analysis by the BWR Owners' Group is incomplete with regards to the modifications which may be required to improve the reliability of Ois instrumentation.
In the interim, additional emergency procedures coupled with existing instrumenta-tion will identify diverse methods of determining inadequate core cooling. We understand that CE, the BWR Owners' Group and the NRC are in communication with regards to this position and that Pilgrim Unit #1 is in compliance with your requirements.
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2.1.4 Containment Isolation Provisions for PWR's and BWR's Position 1.
All containment isolation system designs shall comply with the recommendations of SRP 6.2.4; i.e. that there be diversity in the parameters sensed for the initiation of containment isolation.
2.
All plants shall give careful reconsideration to the definition of essential and non-essential systems, shall identify each system determined to be essential, shall identify each system determined to be non-essential, shall describe the basis for selection of each essential system, shall modify their containment isolation designs accordingly, and shall report the results of the re-evaluation to the NRC.
3.
All non-essential systems shall be automatically isolated by the containment isolation signal.
4.
The design of control systems for automatic containment isolation valves shall be such that resetting the isolation signal will not result in auto-matic reopening of containment isolation valves.
Reopening of containment isolation valves shall require deliberate operator action.
Response
1.
Diversity of parameters sensed for the initiation of containment isolation shall be provided in accordance with SRP-6.2.4.
Diversity in Parameters Sensed for Initiation of Containment Isolation A.
Secondary Containment Isolation (FSAR 5.3.3.3)
Either of two signals will initiate the secondary containment system.
These signals, which indicate a loss-of-coolant accident inside the drywell are high drywell pressure or low reactor water level.
In addition, radiation monitors in the operating (refueling) floor ventil-ation exhaust duct, which indicate a fuel handling accident, can initiate the secondary containment system.
Secondary containment can also be initiated manually from the control room.
B.
Table 1 summarizes the isolation signal codes (asterisk items only) used by the Primary Containment and Reactor Vessel Isolation System. Addit-ional details may be found in PNPS 1 FSAR Section 7 3.4.7 (Isolation Functions and Settings)
Exceptions to the diverse isolation signals criteria have been identified to the NRC in response to IE Bulletin 79-08.
The NRC has accepted the existing methods for isolation of all valves except the reactor water sample valves, the MSIV drains, and the RWCU supply and return valves.
The reactor water sample valves presently receive only one isolation sig-nal (low-low reactor water level) that meets the diverse isolation criteria. A second isolation signal containing high drywell pressure will be added to the existing logics to provide the diverse signals re-quired as these valves have no effect on plant safety.
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Changing the isolation signals to the MSIV drains, however, could affect plant safety. Operation of these valves to more restrictive isolating requirements than the MSIV's could possibly result in condensate buildup between the MSIV's thus preventing operation (opening) of the MSIV's or damaging the steamlines to the condenser. Either failure will needlessly eliminate the condenser as a heat sink after a unit scram thus removing one possible method of cool down.
RWCU suction and return line isolation valves are currently provided with only one containment isolation signal in addition to the process isolation signals. The RWCU system intentionally remales active to keep cleansing the vessel water during the situation where h;gh drywell pressure exists because the drywell coolers are not operating or a small break LOCA occurs.
The small break LOCA could also result in a high drywell pressure condition without reaching a low reactor vessel level condition. It is desirable to keep the RWCU operating under these conditions.
Response
2.
Definition of Essential and Non-Essential Systems A.
Source of Definition is NUREG 0578, Pg. A-14 B.
Definitions:
1.
Essential Systems: Those systems that should be selectively isolated during containment isolations only after it is estab-lished that the use of these systems will not be needed for an
. accident or abnormal transients.
2.
Non-Essential Systems: Those systems not needed for mitigation of an accident or abnormal transient and which should be immediately isolated during containment isolation.
C.
FSAR (Sect. 1.5.2.6.2)
Criteria for implementing Definitions:
1.
A primary containment shall be provided to completely enclose the reactor vessel.
It shall be designed to act as a radioactive mat-erial barrier during or following accidents that release radio-active material into the primary containment.
It shall be possible to test the primary containment integrity and leak tightness at periodic intervals.
2.
A secondary containment that completely encloses both primary contain-ment and fuel storage areas shall be provided and shall be designed to act as a radioactive material barrier.
3.
The primary and secondary containments, in conjunction with other engineered safeguards, shall act to prevent the release of radioactive material from the containment volumes from exceeding the guideline values of applicable regulations.
4.
Provisions shall be made for the removal of energy from within the primary containment as necessary to maintain the integrity of the containment system following accidents that release energy to the primary containment.
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5.
Piping that both penetrates the primary containment structure and could serve as a path for the uncontrolled release of radioactive material to the environs shall be automatically isolated whenever such uncontrolled radioactive material release is threatened. Such isolation shall be effected in time to prevent radiological effects from exceeding the guideline values of applicable regulations.
Classification of Systems Essential Systems a.
RHR (except head spray) b.
RCIC d.
Core Spray (except test lines) e.
HPCI f.
Main Steam Flow Instrumentation g.
Drywell Pressure Instrumentation h.
RBCCW - see note
- i. Containment Atmospheric Control System Non-Essential Systems a.
Main Sters b.
Feedsater c.
Reactor Water Sample d.
Control Rod Drive Hydraulic Return e.
Control Rod Drtie Inlet and Outlet f.
RHR Reacte. Head Spray g.
Rea. cor Water Cleanup h.
Core Spray Test Line to Suppression Pool
- i. Drywell Equipment Drain J.
Drywell Floor Drain k.
Travsrsing In-core Probe l
1.
Service Air m.
Instrument Air Note: RBCCW has 2 Class C containment isolation valves (check valves and motor operated gate valve), one valve per containment penetration.
Class C valves are on process lines that penetrate the primary contain-ment but do not communicate directly with the reactor vessel, with the l
primary containment free space, or with the environs. Class C lines require only one valve which closes automatically by process action (i.e reverse flow) or by remote manual operation from the control room.
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Response
3.
All Non-Essential Systems Shall be Automatically Isolated by the Containment Isolation Signal Table 1 gives a listing of all non-essential systems and their respective isolation signals.
Items which require greater detail are described in the next few paragraphs:
Tip Valves Section 5.2.3.5.2 of FSAR Tip system guide tubes are provided with an isolation valve which closes automatically upon receipt of proper signal and after the TIP cable and fission chamber have been retracted.
In series with this isolation valve, an additional or backup isolation shear valve is included. Both valves are located outside the drywell. The function of the shear valve is to assure integrity of the containment in the unlikely event that the other isolation valve should fail to close or the chamber drive cable should fail to retract if it should be extended in the guide tube during the time that containment isolation is required. This valve is designed to shear. the cable and seal the guide tube upon an actuation signal. Valve position (full open or full closed) of the automatic closing valves will be indicated in the control room.
Each shear valve will be operated independently.
The valve is an explosive type valve and each actuating circuit is monitored.
In the event of a con-tainment isolation signal, the TIP system receives a command to retract the traveling probes. Upon full retraction, the isolation valves are then closed automatically.
If a traveling probe were jammed in the tube run such that it could not be retracted, instruments would supply this information to the operator, who would in turn investigate to determine if the shear valve should be operated.
Section 7.5.9.2.2. of PNPS 1, FSAR A valve system is provided with a valve on each guide tube entering the primary containment. These valves are closed except when the TIP subsystem is in operation. A ball valve and a cable shearing valve are mounted in the guide tubing just outside of the primary containment.
They prevent the loss of reactor coolant in the event a guide tube ruptures inside the reactor vessel.
A valve is also provided for a gas purge line to the indexing mechanisms.
A guide tube ball valva opens only when the TIP is being inserted. The shear valve is used only if a leak occurs when the TIP is beyond the ball valve and power to the TIPS fails. The shear valve, which is controlled by a manually operated protected switch, can cut the cable and close off the guide tube.
The shear valves are actuated by detonation squibs. The continuity of the squib circuits is monitored by front panel indicator lights in the control room.
A guide tube ball valve is normally de-energized and in the closed position.
When the TIP starts forward the valve is energized and opens. As it opens it l
actuates a set of contracts which gives a signal light indication at the TIPS control panel and bypasses an inhibit limit switch which automatically stops TIP motion if the ball valve does not open on command.
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Compressed Air System Valves Section 10.11.3.1 of PNPS 1 FSAR Pressure loss in the high pressure system, sensed by several pressure switches, will cause valves in the service air header, the low pressure service air cross-around line, and the non-essential instrument air header to close in a cascading sequence thus leaving the essential instrument air header as the only header drawing air from the receivers in the event that supply pressure decreases.
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Response
4.
Reopening of Containment Isolation Valves During meetings with the NRC on December 11, 1979 and March 18, 1980, the NRC accepted the PNPS reset circuits for the Balance of Plant Isolation valves with the exception of the primary containment vent and purge system. To reset the BOP isolation logic af ter a scram, it is first necessary to reset the General Electric isolation logic at panel C905 and then reset the BOP logics at panel C7.
The only change required by the NRC to this reset function is the replacement of the existing reset pushbuttons on panel C7 with keylocked selector switches, and this has been done.
The control circuits for the primary containment vent and purge system iso-lation vlaves have been revised by wiring valve control switch contacts (either directly or through auxiliary relays) parallel to the normally closed reset selector switch contacts. The control switch contacts, closed whenever a control switch is in an open position, provide a path for maintaining the trip relays energized for isolation (independent of reset switch position) until all control switches are moved to close. At that time,the isolation logics can be reset by operation of the keylocked reset selector switch.
Also affected by NUREG 0578 were the MSIV's, the reactor water sample valves, the drywell sump effluent valves, and the " emergency open" position of the vent and N2 makeup valves.
The MSIV's control circuits were revised by wiring "close" contacts from each MSIV switch in series with the applicable trip logic reset contacts. This arrangement requires the operator to move all MSiv control switches to "close" before the trip logics can be reset after an automatic isolation.
The control circuits for the " emergency open" position of the vent and N2 makeup valves have been revised by wiring contacts from the applicable control switches (via auxiliary relays) in series across the trip relay sealin contact.
This arrangement, similar to that used on the MSIV's, requires that all valve control switches be closed before the trip relay is reset.,In addition, we are replacing the existing control switches as recommended by the NRC, to allow operation between the "Open" and "Close" without a key and require a key to get into the " Emergency Open" position.
The control circuits for the remaining valves with reset problems have been modified by replacing maintained contact control swiches with three position, spring return to normal switches. These switches used in conjunction with auxiliary relays provide sealin circuits which,when tripped, will require operator action to open the valves after the isolation logics have been reset.
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TABLE I ISOLATION SIGNALS TO NON-ESSENTI AL SYSTEMS System Penetratlon Normal Status (l)
Isolation Slanal la Main Steam Lines Open B, C, D, P, Q b Main Steam Drains Closed (2)
B,C,D,P,Q r
2 Reactor Feedwater Open Rev. Flow (Check Valves) 3 Reactor Water Sample Closed (2)
B,C,D,P,Q,Fjf 4 CRO Return Note 4 Rev. Flow (Check Valves) 5 CRD in and Outlet Note 4 Note 4 6 RHR Head Spray Closed AUF
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7 Reactor Water Cleanup Open A, W, Y, J, RM 8 CS Test Line Closed (2)
G 9 Orywell Equip. Drains Open B, F 10 Drfwell Floor Drains Open B',
F lla TIP Primary Closed (2)(3)
F, A b
Backup Open RM (Explosive Shear Valve) 12 Service Air Closed Rev. Flow (Check Valve) 13 instrument Air Open inside - Rev. Flow (Check Valve)
Open Outside - RM NOTES:
(I)
Normal status position of a valve is the position during normal power operation of the reactor.
(2)
Valve can be opened or closed by remote manual switch for operating convenience during any mode of reactor operation except when automatic signal is present.
(3)
Signal "A" or "F" causes automatic withdrawal of TIP probe, then valve automatically closes by mechanical action.
(4)
CRD solenoid valves are normally closed, but they open on rod movement and during scram.
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.7 ISOLATION SIGNAL CODES FOR TABLE I SIGNAL DESCRIPTION A*
Reactor vessel low water level - scram and close isolation valves except main steam lines.
B*
Reactor vessel low low water level - Initiate RCIC, HPCI and close main steam line Isolation valves.
C*
High radiation - main steam line (also causes scram).
D*
Line break - main steam line (steam line high space temperature or high steam flow).
E Reactor low low icvel or nigh drywell pressure - select LPCI and close other loop valves.
F*
High drywell pressure - close RHR/ shutdown cooling and head spray plus the RHR to radweste valves.
G Reactor vessel low water level and low pressure; or high drywell pressure -
Initiate Core Spray and RHR systems.
J' line break in cleanup system - high space temperature, or high flow.
K' Line break in RCIC system steam line to turbine (high steam line space temperature or high steam flow) or low steam pressure.
L' Line break in HPCI system steam line to turbine (high steam line space temperature or high steam flow) or low steam line pressure.
M*
Line break in RHR shutdown and head cooling (high space temperature; alarm only; no auto closure).
P*
Low main steam line pressure at inlet to main turbine (RUN mode only).
S Low drywell pressure - close containment spray valves.
T Low reactor pressure permissive to open core spray and RHR-LPCI valves.
U High reactor vessel pressure - ciose RHR shutdown cooling valves and head cooling valves.
W High temperature at outlet of cleanup system nonregenerative heat excha nger.
Y Standby liquid control system actuated.
l RM" Remote manual switch frem control room.
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Q Reactor high water level - Isolate main steam !!ne (except in run mode).
X RCIC or HPCI steam supply valve (as applicable) not fully closed.
'These are the isolation functions of the primary containment and reactor vessel e
isolation control system; othsr functions are given for Information only.
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.--.,...n 2.1.5.a Dedicated Penetrations for External Recombiners or Post-Accident Purge Systems In order to meet the requirements of 2.1.5.a for Pilgrim Unit 1, a subsystem to the' containment vent and purge system has been added for Post Accident Containment Combustible Gas Control (CCGC). This will provide combustible gas control after a design basis LOCA with loss of offsite power and a single active failure without requiring access to areas that may not be reasonably accessible.
To meet the above requirements, 16 solenoid valves have been added and arranged to provide redsndant oaths to and from the drywell and torus for venting and N2 makeup. The solenoid valves are designed to remain closed against maximum con-tainment pressure, to vent containment so that the maximum containment pressure will not be exceeded, and to provide a nitrogen flow sufficient to maintain the oxygen concentration inside containment below the flammability limits.
The valves in redundant paths are powered from independent Class IE distribution systems each of which is powered from an emergency diesel generator af ter a loss of offsite power. The control switches for redundant valves are located in sep-arate Class IE control panels in the main control room. All equipment and conduit is located in seismically designed, missile protected buildings, except all the fill connections which are located outside of secondary containment but separated.
Redundant conduit systems are separated commensurate with identified hazards.
All conduit and equipment is supported to meet seismic Category I requirements.
The solenoid valves are ASME III Class 2 and are qualified environmentally and seismically to the requirements of IEEE 323-1974, IEEE 382-1973, and IEEE 344-1975 for the expected conditions. The valves are rated at 120VAC and are designed to operate between 80 and 110 percent of rated voltage.
This range is compatible
- with expected bus voltages at PNPS.
The control switches have been qualified to the requirements of IEEE 323-1974 for operation in a control room environment. The switches are mounted on Class IE panels and the combination has been qualified to IEEE 344-1975 for the design basis earthquake. The switch's electrical ratings exceed loading requirements.
The cable and wire used for this modification have been qualified to IEEE 383-1974 for fire and ambient conditions exceeding those required for this installation.
Ihe 600 volt #12 AWG control cable has voltage and current capabilities well above that required.
Based on the above, it can be seen that this design will perform its required func-tion assuming a loss of of fsite power and a single failure and that the installation is designed to withstand the effects of the most severe natural phenomena postulated.
Control of the solenoid valves is remote manual, there is no automatic isolation capability.
Isolation signals have not been provided because:
- 1) The valves are always keylocked closed during normal operation.
- 2) The valves are required to be operated during a high drywell pressure condition and must be available independent of reactor water level. High drywell pressure and low reactor water level are the normal containment isolation signals.
Indicator lights are provided to continuously monitor valve position. The indicators are driven by reed type limit switches mounted within the valve electrical housing.
Contacts from all control switches are wired to an annunciator window to provide an alarm when a valve is open.
The only interf ace between the new components and existing systems are that the new valves are piped in parallel to the existing containment vent and purge valves, the new valves discharge to the standby gas treatment system, and both the new and existing vent / purge valves are powered from different circuit breakers in the same distribution panelboards. The standby gas treatment system can process any flow through the added valves. All containment vent and purge valves (both new and existing) receive power from distribution panelboards Y3 and Y4.
Each panel is powered through a 3 kVA, 480-120 volt transformer. The maximum load on this transformer occurs af ter a LOCA when all the new solenoid valves could be energized at the same time. The total loading due to operation of all components required af ter a LOCA does not exceed the transformer rated capacity. Additionally dedicated primary containment penetrations are being provided for external recom-biners.
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2.1.6.a Integrity of Systems outside Containment Likely to Contain Radio-active Materials I.
Position A program has been implemented to reduce leakage from systems outside contain-ment that could contain highly radioactive fluids during a serious transient or accident.
The program consists of a quarterly surveillance during system operation to identify leakage from liquid systems, and the pressurized portions of gaseous systems. Leakage rates will be estimated and corrective action taken to maintain leakage as low as possible.
II.
Systems Included in Program 1.
HPCI 2.
RCIC 3.
Core Spray 4.
RHR 5.
Reactor Water Cleanup (Let Down Portion) 6.
SBGTS (Discharge only) 7.
Radwaste Collection System from Reactor Building 8.
Sample System (from Recire and RWCU only)
III. Systems Not Included in Program 1.
Main Steam downstream of Outboard MSIV's 2.
Turbine Extraction Steam and' Drains 3.
Condensate and Demineralized Water Storage and Transfer 4.
Condensate and Feedwater 5.
Off Gas 6.
Seawater 7.
Salt Service Water 8.
RBCCW 9.
TBCCW i
10.
Fire Protection 11.
Service and Instrument Air 12.
Makeup Demineralizer
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Fuel Pool Cooling and Demin.
14.
Clean Radwaste 15.
- 16.
CRD 17.
Recire.
IV.
Method of Inspection A.
Steam.and Liquid Systems Systems will be operating or pressurized to their operating pressure.
All accessible pipe runs will be inspected for evidence of leakage from:
1.
Mechanical joints - bolted flanges and pump casings 2.
Pump shaft seals 3.
Valve packing and bonnet gaskets 4.
Relief Valves 5.
Rupture discs.
All leakage will be identified with leakage rates estimated based on visual observation and frequency and duration of sump pump operation.
B.
Gaseous Systems The positive preseure (greater than atmospheric) portions of gaseous systems will be inspected by injecting Freon into the low pressure end of the system and using a Freon detection device to detect leakage at the following areas:
1.
Mechanical joints - bolted flanges 2.
Blower shaft seals 3.
Valve packing All leakage will be identified with leakage rates estimated. A liquid
" snoop" solution will be used to augment the inspection.
V.
Repairs All leaks identified during surveillance inspections will be evaluated through the normal maintenance process with repairs scheduled to reduce leakage to as low as practical.
Boston Edison's initial inspection under this program identified the following leakage points:
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- 1) RCIC - 5 GPM thru rupture disc.
- 2) RHR
.1 GPM thru MOV packing
- 3) RWCU
.1 GPM thru hand valve packing
.5 GPM thru HE drain valve seat 1 GPM thru A0V plug cap To date, all identified leakage points have had work orders issued against them and will be reduced to as low as practical.
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2.1.6.b Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Which may be used in Post-Accident Operations I.
Introduction In compliance with the position described in NUREG-0578 as modified by the September 13, 1979 and October 30, 1979 NRC letters to all power reactor licensees,a radiation and shielding design review of Pilgrim Nuclear Power Station Unit #1 (PNPS) was conducted. This review was directed towards identifying the location of vital areas and equipment in spaces around systems that may, as a result of an accident, contain highly radioactive materials. The objective of this review was to determine areas where personnel occupancy may be unduly limited and safety equipment unduly de-graded by the radiation fields during post-accident operations of these systems. The guidelines for personnel accessibility were those provided in the above-mentioned correspondence. For safety equipment operability avail-able equipment qualification specifications were referenced.
II.
Assumptions
- a. Accessibility Criteria Personnel accessibility post-accident was reviewed both for areas which required continuous occupancy and for other areas which were ident-ified as requiring infrequent access.
For all areas of the plant this shielding review extended the occupancy considerations up through and beyond the initial cold shutdown condition.
For areas requiring infrequene access, General Design Criteria 19 was applied (i.e. as defined by the October 30, 1979 NRC letter).
In extending the analysis to the cold shutdown condition under the con-servative source term assumptions, the analysis considered anticipated operator exposures in the performance of expected maintenance.
Un-attended run times were also estimated based on documented experience,
- b. System Requirements Cold Shutdown and long-term core cooling requirements necessitated identification of systems which would be required to properly function.
These systems were separated into two groups for the purposes of eval-uating equipment operability and personnel accessibility in the presence of a radition field resulting from an accident of the assumed magnitude:
Group I Systems required to properly function to achieve low pressure core cooling.
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Group II - Systems required to properly function to maintain long-term l
core cooling.
System categorization is provided below:
Group I Reactor Protection System High Pressure Coolant Injection System Automatic Depressurization System RHR System (LPCI Mode)
Group II Primary Containment Isolation System RHR System (Shutdown Cooling and Suppression Pool Cooling Modes)
Core Spray System Standby Gas Treatment System Primary Containment Integrity 1
Secondary Containment Integrity Emergency AC Power System t
Emergency DC Power System Reactor Building Closed Cooling Water System Salt Service Water System Contro,1 Room Environmental Systems Onsite Technical Support Center Environmental Systems Containment Atmosphere Control System A shielding review of systems required for prompt response and high pressure core cooling and reactor water level control was not conducted.
The basis for this action was that upon recognition of an accident with substantial fuel failure the plant would be promptly brought to a low pressure state and placed in a low pressure cooling mode.
High pressure systems would not be required for extended operation nor would subsequent access to these systems be required.
- c. Source Terms The shielding review was based on photon spectrum developed by "0RIGEN-The ORNL Isotope Generation and Depletion Code," ORNL-5628, M.J. Bell.
The code assumed a 1998 Mwt. core with 3 years of operation at 100% power.
The resulting spectrum was used to analyze the maximum accident dose rates in piping containing radioactive fluids and gases following an accident of the magnitude prescribed for this review.
In developing an appropriate source term, Regulatory Guides 1.3,1.7, and Standard Review Plan 15.6.5 were consulted as provided in the October 30, 1979 NRC letter.
In reviewing these assumptions, it was decided to initially increase the assumed activity in the liquid containing systems considered. This was done in order to reflect a more conservative source term in the unlikely event of an accident of this magnitude.
For liquid containment systems, this resulted in a source term consisting of 100%
of the core's content of iodine and bromine to be instantly released to
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the reactor coolant. To reflect possible cesium leachout, 100% of the core's cont'ent of cesium was also assumed to be instantly released as well.
In addition, 1% of the core's particulate inventory was also assumed to be released.to the reactor coolant at the same time.
For gaseous systems, the assumptions used for the source term were those specified by the NRC, i.e., 100% of the core's content of noble gases and 25% of the core's content of the iodine be instantly released to the reactor coolant.
In performing this review reference was initially made to the PNPS FSAR Chapter 14.
While Chapter 14 was directed towards calculating LOCA off-site doses, the analyses performed do provide useful information in that comparable source. terms were considered. A subsequent review indicated that more conservative assumptions than those used in the FSAR could be made and that, on the basis of engineering judgment, the more limiting assumptions were more appropriate to initiate the shielding review.
Con-sequently, the most limiting assumptions for activity release to the reactor coolant were made and the corresponding doses calculated.
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a III. Plant Accessibility Assessment and Recommended Notifications Reactor Building A review of personnel accessibility to the Reactor Building indicates that entry to most areas will be practically precluded'for the first thirty days following the postulated accident due to high radiation fields. Maintenance during this period on plant systems necessary for cold shutdown would be severely limited under present conditions.
Due to the significant radiation levels from the major core spray discharge piping, shielding would improve accessibility to the Reactor Building at elevations 23' and 51'.
This would afford more immediate access to general areas of the Reactor Building in the event of an accident of a magnitude less than that analyzed. Analysis is in progress to determine if additional shielding can be designed to be within seismic structural plant capability.
t In addition to the core spray piping, another area of concern is the Reactor Building Truck Lock Door. This access provides a double barrier to the plant yard and an air-lock feature of the secondary containment. As access through this door would not be necessary subsequent to an accident, closure of the inner door would provide additional reduction in radiation levels in the yard with no loss of operability by eliminating the gaseous source term within the truck lock space. Currently, this door must be operated locally and therefore may not be closeable following a major accident. PNPS will incorporate a remote and accessible door closure capability.
In the event of a postulated single failure, operator action to position cer-tain valves would be necessary in order to maintain the analyzed combustible gas concentration inside containment to less than explosive levels. Due to the high radiation levels in the vicinity of these valves, modification of the valves and the control systems is being made to obviate the requirement for operator accessibility.
Control Room The control room is available to continuous occupancy based on an average exposure of less than 15 mr/hr as provided by the October 30, 1979, NRC letter.
Shielding between a common wall section between the Reactor Building and the Control Room will be subjected to further detailed analysis to determine if shielding additions or other modifications are required.
Diesel Generator Rooms The Diesel Generator Rooms radiation exposure is not expected to exceed the limits of GDC 19 in support of expected operator actions following the postulated accident.
Turbine Building & Radwaste Areas Following an accident of the magnitude postulated for this review, PNPS would conduct operations necessary to achieve cold shutdown and long-term core cooling in a manner which limits the spread of contamination.
This operational phil-osophy would maximize the use of post-accident-backup systems to limit the contamination of installed equipment.
This includes prompt isolation of con-tainment and other sumps which could inadvertantly transfer contaminated water m
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t from the primary systems outside containment. The Turbine Building and systems contained therein are not expected to be needed for post-accident operations in support of achieving cold shutdown or long-tcrm core cooling and should be isolated following a major accident with substantial fuel failure.
On that basis, access to the Turbine Building or radwaste areas is not con-sidered necessary nor would radiation induced equipment failures be con-sidered limiting. No hardware modifications are recommended in this area.
Residual Heat Removal Rooms The RHR Rooms (and other quadrants) are expected to be subjected to high radiation' levels following an accident of the magnitude postulated for this review. Accessibility for equipment maintenance in the RER Rooms could be precluded for at least 1 year following this accident assuming the maximum spread of contamination. This is due to the anticipated high coolant activities, the tight quarters and the large physical size of piping with contaminants within these rooms.
Accident procedures will be developed for the Residual Heat Removal System operation to maximize dilution to reduce in-plant dose consequences.
Specifically, the mixing of suppression pool water (plus additions from the CST) with the reactor coolant would result in a substantial dose reduction from systems containing reactor coolant through a dilution effect, following an accident.
In addition to the procedure modification, PNPS is evaluating the capability for future installation of equipment necessary to support long-term core cooling and reactor water cleanup following a major accident. This capability will obviate the necessity for contaminating the installed Liquid Waste Management System and environs.by attempting to process liquids contaminated to levels in excess of the system's design capability.
This installation would consists of tie-in lines and isolation valves to the RHR system at locations and in a manner which would allow for subsequent adaptation of shielded demineralizer/ filtration and cooling systems as may be required for post accident recovery. These tie-ins could terminate in an accessible l
area, such as the Radwaste Truck Lock.
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of a backup clean-up and heat removal system, the following activities are in planning or being implemented:
(1) Determination of the current reliability for prolonged, unattended operation of the long-term core heat removal systems. This effort will involve both deterministic and probabilistic assessments of safety equip-ment qualifications and equipment operability.
Probabilistic methods will be applied to assess the current and future l
documented experience of component operability using established stat-l istical procedures. Preliminary indications are that the components in question have demonstrated high reliability and that the RER function will be available unattended for sufficient time-to allow post-accident backup systems to function.
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(2) Provide shielding to protect essential instrumentation and control systems to within their qualified limits from the deleterious effects of high radiation fields.
Based on the results obtained from the reliability assessment, sufficient local shielding will be added.
This effort will be incorporated into the program in progress to assess electrical equip-ment qualifications associated with IE Bulletin 79-01.
(3) Ensure that the preventative maintenance program maximizes current.
system reliability to perform the RHR function during prolonged, unattended operation. This effort should be tailored to each component's maintenance history and the manufacturer's recommendations and should encompass all safety-related systems including systems necessary for cold shutdown and long-term core cooling.
(4) Add the capability to allow for the future installation of equipment necessary to support long-term core cooling and reactor water cleanup.
As discussed previously, this capability will minimize the spread of contamination to the existing systems at blanked-off taps. Piping pen-etrating containment leading to these taps will be sufficiently shielded to allow necessary personnel access so as not to exceed GDC 19.
IV.
Conclusions This iesign review assessed the implications of a postulated accident of a serious magnitude. While an accident of this magnitude is extremely unlikely, PNPS assumed a source term more conservative than dicteted in assessing the implications.
PNPS was specifically designed to mitigate major design basis events with no access outside the control room necessary. PNPS will add to this existing capability through procedure modifications, shielding and backup systems so as to minimize contamination and operator exposure and maximize the operational flexibility availability to the operating staff following an accident.
Area radiation assessments resulting from this review will be utilized in developing post-accident procedures and appropriate modifications. Additional refinements will be performed in specific instances which warrant further analysis.
During the March 18, 1980 NRC/BECo meeting on TMI modifications at Pilgrim Nuclear Power Station, the NRC requested that Boston Edison include dose rate maps with our submittal.
In response to that request, those maps are attached.
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2.1.8.a Post Accident Sampling System We have performed a design and operational review of the existing reactor coolant and containment atmospheric sampling systems at Pilgrim 1 We have found that the present sampling system may be inaccessible under design conditions and that the present reactor coolant sample points may tap into stagnant areas that are not representative of the bulk system parameters. To correct these deficiencies, we intend to provide a new sampling station in an accessible location and to add new reactor coolant sample points. The new sample station will have provisions to protect personnel from both airborne and line source radiation hazards so that sampling operations will not result in personnel doses exceeding 3 rem whole body and 18 3/4 rem' extremities.
In addition, gaseous eff!uents from sampling operations will be controlled to insure that public health and safety are not impacted. This is especially important since we must locate the new sample station outside of the Reactor Building to insure personnel access. The nev liquid sample points will be connected to the sensing lines to the flow transmitters of jet pumps 5 and 15 (pressurized sample) and to the discharge of the RHR pumps on both RHR loops (de-pressurized sample).
These locations will insure representative samples since they will not be stagnant after an accident. Gas samples from the torus (2 samples) and Drywell (4 samples) will also be available.
Liquid samples will be analyzed for dissolved hydrogen and isotopically. Cas samples will be analyzed for hydrogen, oxygen, and isotopically.
Safety grade components will be used up to and including the second automatically operated isolation valve in each sample line. The balance of the system will consist of non-safety grade components and materials.
Sample lines will be redund-ant up to and including the second isolation valve in each sample line. Interim measures have been implemented to improve the capability of the existing sample station. Equipment is available to reduce the airborne radiation hazards assoc-lated with drawing a liquid sample with the present sampling system.
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2.1.8.b Noble Cas Effluent Monitor BECo is providing a systen of noble gas effluent monitors. The noble gas system consists of three field mounted detectors, (ion chambers), tied to a module mounted in the control room.
The three effluent points being monitored are:
1.
Reactor Building Vent 2.
Main Plant Stack 3.
Turbine Building Exhaust The detectors are provided with range of 10-1 to 104 R/hr. The hardware provided constitutes a single channel of a non-safety related system which is powered from an emergency power source.
The main control room module provides signals to three indicators and a three pen recorder giving effluent radiation levels at the three locations stated above.
1 Calculations have been performed to relate monitor readings to effluent release rates. These correlations have been incorporated in PNPS Procedures, i
2.1.8.c Improved In-Plant Iodine Instrumentation The intent of this position has been met by successfully completing the following:
- 1. ' Silver Xeolite Cartridges have been purchased and are stored on-site.
2.
Cartridge counting equipment has been procur and installed.
3.
Radionuclide standards for System Calibration have been prepared.
4 Procedures will be modified and Health Physics Training completed prior to plant restart.
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2.1.9 Accident Analysis Work by GE and the BWR Owner's Group is ongoing concerning this position.
Dev-elopment of emergency procedure guidelines, analysis of FSAR Section 15 events, eme'rgency procedure revisions and operator retraining are proceeding in accord-ance with a schedule established with the Bulletins and Orders Task Force.
Procedures which have undergone revision after Position 2.1.9 re-analysis to date include:
5.4.2. P,ipe Break Inside Primary Containment 5.4.4 Pipe Break Inside Reactor Buf1 ding 5.4.5 Pipe Break Outside Secondary Containment l
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2.2.1.a Watch Engineer Responsibilities The responsibilities of the Nuclear Watch Engineer were emphasized in NOD Directive 80-1.
The directive will be reissued periodically to assure the primary management responsibilities of the Nuclear Watch Engineer are under-stood. Pilgrim Nuclear Power Station Procedure 1.1.1 has been revised to reduce the administrative duties and to improve the functional performance of this position.
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I 2.2.1.b Shift Technical Advisor The Boston Edison Company has created the position of Shift Technical Advisor (STA) at Pilgrim Station Unit #1.
This position is accountable for the safe operation of Pilgrim Unit #1 through contributing to assessment of plant con-ditions during normal operation and transients consistent with technical specifications, procedures and regulatory requirements. The Shift Technical Advisor Group is independent of Operations and maintains direct line reporting to the Assistant Station Manager. The Shift Technical Advisor is assigned to a specific shift and there are three (3) shifts per day. One (1) STA is on duty at all times during normal plant operation. The STA maintains a high awareness of adfety in plant operation and provides analysis and reporting of plant conditions during operation and transients. The STA provides technical expertise to the Watch Engineer in order to help the Watch Engineer recognize, diagnose and respond to unusual events.
The STA provides the perspective and the time for assessing plant conditions by independently monitoring plant safety.
On a daily basis, the STA communicates with other STA's to report ongoing plant conditions and the status of any special circumstances. Also on a daily basis, the incumbent (s) interface with the Watch Engineer (s) to assess the effectiveness of operations. On a frequent basis (weekly) the incumbent (s) receive update reports of unusual occurrences from the Reliability and Safety Assessment Group.
As required, the incumbent (s) provide technical comments and recommendations on plant operations and responses of operators to transients for the Chief Operating Engineer. The STA provides assessment of transients or conditions during an event to the On-Site Technical Support Center Staff.
The role of the incumbent is essentially one of monitoring plant conditions and maintaining a close inter-face with the Watch Engineer regarding plant operations. The scope of the pos-ition also includes providing recommendations to the Watch Engineer as requested for actions to be taken by the Nuclear Plant Operators in response to an unusual event or transient.
Boston Edison believes that our program meets the intent of NUREG-0578 Position 2.2.1.b, Shift Technical Advisor.
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2.2.1.c Shift Turn over Pilgrim Nuclear Power Station Procedure 1.3.18 has been revised to assure the continuous transmittal of operational information and status of systems essential to the prevention and mitigation of an accident or transient.
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_ Control Room Access Pilgrim Nuclear Power Station has developed and implemented Procedure 1.3.32 which establishes authority and responsibility of the person in charge in Control Room to limit access and to establish a clear line of authority in the. Control Room during an emergency.
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2.2.2.b Technical Support Center The training room in the lower level of the guardhouse has been designated as the short-term and interim technical support center. The details presented in Reference (D) remain valid with the following:
2.c The existing structure affords a protection factor of three for external gamma radiation. A protection factor of one is afforded for airborne particulates, iodine and noble gases.
3.b The existing air intake is filtered for suspended particles.
3.c The TSC will be monitored for airborne particulate and direct radiation using portable equipment.
5 Dedicated two-way communications are provided between the TSC and the Control Room and the TSC and the Emergency Control Center.
6 Selected documents and drawings are available in the records center which is in close proximity to the TSC.
The latest revision of the emergency plan defines the role and location of the technical support center.
During a site visit by Mr. bave Verre111 and the TMI Task Force on March 18, 1980, the technical supperi center was visited and details of the design were discussed.
Based on that visit, we understand that our location and design for the technical support center is acceptable to the NRC.
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2.0.2.c Operational Support Center The area used as the lunchroom at PNPS is designated as the operational support center. Comununication with the control room is provided. The draft revision f
to.the Emergency Plan describes the methods and lines of communication of the operational support center.
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