ML20006A263
| ML20006A263 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/15/1990 |
| From: | Bird R BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-REGGD-01.097 BECO-90-010, GL-82-33, TAC-51119, NUDOCS 9001260054 | |
| Download: ML20006A263 (35) | |
Text
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Pilgrim Nuclear Power Station Rocky Hdi Road Plymouth, Massachusetts 02360 i BECo 90- 010 i Ralph G. Bird January 15, 1990 Senior Vice President - Nudear U.S. Nuclear Regulatory Commission ;
Document Control Desk ,
Hashington, DC 20555 License DPR-35 !
Docket 50-293 .
Summary of Compliance with Regulatory Guide 1.97, Revision 3 Concernino Emergency Response Caoability (TAC 51119)
To assist the NRC in its review of compliance with Regulatory Guide 1.97, !
Revision 3, Boston Edison Company is providing the attached summary of "
- compliance for the Pilgrim Nuclear Power Station. This summary restates L compliance information previously submitted to the NRC, provides new information for specific variables, and identifies open items requiring i additional work. All new information in this summary of compliance is ,
! identified with revision bars to aid the reviewer.
l
.B d DMV/amm/3764 l
Attachment:
Summary of Compliance with Regulatory Guide 1.97, Revision 3 for s the Pilgrim Nuclear Power Station ;
cc: Mr. D. Mcdonald, Project Manager Division of Reactor Projects - I/II i Office of Nuclear Reactor Regulation i
' Mail Stop: 1401 U. S. Nuclear Rettlatory Commission '
1 White Flint North 11555 Rockville Pike Rockville, MD 02852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 ;
Senior NRC Resident Inspector Pilgrim Nuclear Power Station ()03 9001260054 900115 l PDR ADOCK 05000293 F PDC
(
- 1 I
, ATTACHMENT TO BECO 90-010 j
SUMMARY
OF COMPLIANCE HITH REGULATORY GUIDE 1.97, R[yISION 3 FOR THE PILGRIM NUCLEAR POWER STATION i
i I.
SUMMARY
A summary of compliance of the post-accident monitoring instrumentation at the L
Pilgrim Nuclear Power Station (PNPS) to the design and qualification criteria of Regulatory Guide 1.97, Revision 3 is provided in Table 1. Compliance information for individual primary containment isolation valves is provided in ,
Table 2. Justifications are provided in Section II for all deviations I identified on these tables. All open items requiring additional work are l identified on these tables with an "0" and are described in more detail in i Section III. 1 i
References are provided for all compliance information that has previously l been submitted to the NRC, All new information in this summary of compliance '
is identified with revision bars.
II. JUSTIFICATIONS FOR DEVIATIONS 1
' A. Drvwell Atmosehere Temnerature (Tvoe A. Category 1 and Tyne D.
Category 2)
The drywell atmosphere temperature instrumentation at PNPS deviates !
- i. from the Regulatory Guide 1.97 recommended range of 40 to 440*F. j Although the drywell atmosphere temperature range of 0 to 400'F at PNPS does not correspond exactly with the Regulatory Guide 1.97 >
g J
recommended range, it does provide sufficient range for monitoring the anticipated design temperature of 281*F, as described in the Final Safety Analysis Report (FSAR). The environmental qualification L bounding drywell temperature for a steam line break inside containment of 330'F and the identified peak temperature of approximately 340'F described in the Emergency Operating Procedures are also adequately covered by the 0 to 400'F range. For this reason, the instrument range at PNPS is acceptable (Reference 3).
B. Containment and Drvwell Hydrogen Concentration (Tvoe A. Category _1 :
and Tvoe C. Cateaorv 1) i This variable deviates from the Regulatory Guide 1.97 recommended range of 0 to 30 percent hydrogen concentration. '
The instrumentation provided at PNPS to measure the concentration of hydrogen in the containment has a range of 0 to 10 percent. This instrumentation was installed at PNPS to meet the requirements of NUREG-0737 Item II.F.1.6, Containment Hydrogen Monitor. As stated in Reference 2, the NRC concluded that the instrumentation provided at PNPS was acceptable as part of their review of NUREG-0737 Item II.F.1.6. Accordingly, the provided instrument range is acceptable.
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C. Coolant Level in Reactor vessel (Tvoe A. Category 1 and Tvon B.
Cateaory 1)
The instrumentation at PNPS to indicate the coolant level in the reactor vessel deviates from the Regulatory Guide 1.97 recommended range of the bottom of the core support plate to the lesser of the top of the vessel or the centerline of the main steamline. At PNPS, i the Regulatory Guide 1.97 recommended range would be from 186 to 604 i inches above the bottom of the vessel. However, the instrumentation ;
provided at PNPS uses two overlapping sets of Category 1 instrumentation to cover the range of 205 to 532 inches. l The instrument range provided at PNPS gives the operator the reactor vessel level indication needed to perform safety functions under both ,
accident and post-accident conditions. These safety functions :
include the automatic and manual actions that may be required to !
restore and maintain reactor vessel water level and to provide core cooling. Level indication below active fuel and greater than the .
high level trip setpoint of ECCS, as recommended by Regulatory Guide i 1.97, does not contribute to information about the accomplishment of ,
plant safety functions for following the course of an accident.
The PNPS reactor vessel water level range is sufficient to keep ,
instruments on scale, utilizing overlapping ranges, at all times when information is required about the accomplishment of plant safety >
functions for following the course of an accident. The existing level indication range at PNPS meets the intent of the .
. recommendations of Regulatory Guide 1.97 (Reference 3). l D. Neutron Flux - APRM. SRM (Tvoe B. Category 1)
Boston Edison has endorsed the BWR Owners' Group's position that a fully-qualified Class IE post-accident neutron monitoring system is not required. The justification for this position is provided in the .
Licensing Topical Report NEDO-31558, March 1988, " Position on NRC Regulatory Guide 1.97, Revision 3, Requirements for Post-Accident i Neutron Monitoring System." The NRC has not yet completed their review of this BWR Owners' Group position on neutron flux monitoring '
and this item remains open (Reference 5).
E. BWR Core Temoerature (Tvoe B. Categorv None and Tvoe C. Category None)
BHR core temperature thermocouples are not provided at PNPS, which deviates from the Regulatory Guide 1.97 recommendation.
BWR core thermocouples would not provide an appropriate diverse indication of water level in the reactor vessel. Specifically,-the thermocouples would not respond for at least 10 minutes following -
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I the uncovering of the core during a small break LOCA. During this period, the reactor operator would receive conflicting information !
frou existing reactor vessel water level indication. Boston Edison '
concludes that in-core thermocouples would not provide the diverse ,
indication of reactor vessel water level described by Regulatory !
Guide 1.97 and they will not be installed at PNPS.
F; Drywell Sumo Level (Tvoe B. Category 3) and Drywell Drain Sumos Level (Tvoe C. Cateaorv 31 Regulatory Guide 1.97 recommends Category 1 instrumentation for these !
variables. The instrumentation provided at PNPS for these variables-is Category 3.
The drywell sumps at PNPS are automatically isolated at the primary y containment penetration should an accident signal occur. For small leaks to the drywell sump, the instrumentation is not expected to experience harsh environments during operation. For larger leaks, the drywell sumps. fill promptly and the sump drain lines isolate due ,
to the increase in drywell pressure, which negates the drywell sump level and drywell drain sumps level instrumentation. In addition, 1 this .nstrumentation neither automatically initiates nor alerts the i operator to initiate operation of a safety-related system in a post-accident situation. Boston Edison concludes that the Category 3
' instrumentation provided at PNPS will provide appropriate monitoring i of the parameters of concern. The NRC concurred with this conclusion i in Reference 2. l i
G. Primary Containment Isolation Valve Positions (Tvoe B. Cateaorv 1) l
- 1. Channel Redundancy Deviations MO 1201-80, Reactor Water Cleanup (RWCU) Return MO 1301-49, Reactor Core Isolation Cooling (RCIC) Pump Discharge NO 2301-8, High Pressure Coolant Injection (HPCI) Pump Discharge A0 5033A, Normal Nitrogen Makeup to Drywell A0 5033C, Normal Nitrogen Makeup to Torus A0 5040A, Torus Vacuum Breaker Isolation Valve ,
A0 50403, Torus Vacuum Breaker Isolation Valve
- CV 5046, Air Supply to the Drywell to Torus Vacuum Breakers ;
1 Each of these primary containment isolation valves are located on Class A or B lines which require two isolation valves in series.
-Check valves, which close on reverse flow, are used in conjunction with the above valves to isolate the lines. Because Regulatory Guide 1.97 specifically excludes check valves from any position indicating requirements, redundant valve position .
indication will not be provided for these lines.
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N0 4002, Reactor Building Closed Cooling Water (RBCCH) Return This primary containment isolation valve is located on a closed cooling water line penetrating the primary containment. It requires only one isolation valve. Position indication for the )
single primary containment isolation valve MO 4002 is provided in i the control room. Boston Edison concludes that single control )
room indication of primary containment isolation valve position i is acceptable for this line. !
- 2. Valves Excluded from Regulatory Guide 1.97 Program
- a. Disarmed Valves MO 1001-60 and MO 1001-63 Residual Heat Removal (RHR) Head l Spray q These valves have been electrically disarmed in the closed l position and do not require valve position indication in the ;
control room to verify primary containment isolation. For this reason, these valves are excluded from the Regulatory j Guide 1.97 program. l
- b. Control Rod Drive (CRD) Direction 31 Control Valves !
FCV 302-120 and -123, CRD Insert .
SV 302-121 and -122, CRD Hithdraw !
These 580 directional control valves, when energized and 1 opened in coordinated pairs, facilitate rod movement either
-f in the insert or withdrawal modes. These valves are normally closed, except during rod movement in normal operation. No position indication is provided for these valves in the control room and they do not receive an automatic primary containment isolation signal (Reference 1). ;
Because these valves are not used to achieve a scram and are ;
not used in a post-accident situation, no position indication -
is required. These valves are excluded from the Regulatory Guide 1.97 program. The NRC concurred with this position in Reference 2. ,
- c. Lines That Terminate Below Sunoression Pool ,
MO 1001-36A and B, RHR Test Return MO 1001-18A and B, RHR Minimum Flow H0 1301-25 RCIC Pump Suction from Torus M0 3301-36, HPCI Pump Suction from Torus NO 1001-7A through -70 RHR Pump Suction MO 1400-3A and B, Core Spray Suction t
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. I L. j These primary containment isolation valves are located on !
lines that terminate below the water level of the suppression !
pool during both normal and accident conditions. No path for gaseous leakage from the containment exists. The position indication of these valves provides no additional information to the operator on the accomplishment of containment i isolation. Therefore, these valves are excluded from the Regulatory Guide 1.97 program. {
- d. Residual Heat Removal (RHR) Discharae to Radwaste i
MO 1001-21 and MO 1001-32 1
These valves are located upstream of the primary containment i isolation valves on the RHR injection line and, therefore, i are not relied upon to perform primary containment isolation. However, these valves do receive a primary i containment isolation signal to ensure proper valve I positioning. These valves.are not containment isolation j valves and they are excluded from the Regulatory Guide 1.97 1 program. '
- 3. Transversing Incore Probe (TIP) Shear and Ball Valves 736A, 736B, 736C 736D.
737A 7378, 737C, 737D ;
Regulatory Guide 1.97 recommends Category 1 instrumentation for
'the position indication of these primary containment isolation
-valves. Category 3 position indication is provided for these .)
valves at PNPS. ll The TIP primary containment isolation design is commensurate with ;
the importance to safety of isolating that system, and has been I previously reviewed and accepted by the NRC on numerous dockets.
The TIP guide tubes are normally closed by the TIP ball valves.
A TIP scan requires insertion of.the TIP probes into the reactor vessel for a period of approximately four hours per month. Over ;
a one-year period, this amounts to less than 2% of the time the '
plant is operational. In the event of a LOCA, the TIP sp tem design will reliably provide automatic isolation of any open TIP :
guide tubes by providing automatic retraction of the TIP cable 1 L. followed by automatic closure of the TIP ball valves. Only in the case that the ball valve fails to automatically close, the shear valve is manually actuated by detonation squibs. However, ,
because the TIP system electrical circuits are not safety grade l and not separated, failure to isolate TIP guide tubes could be .
postulated.
The most likely sequence of events leading to fission product release through the TIP guide tubes has a probability of occurrence of about 5 X 10E-13 per reactor year. Using extremely i J
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conservative Regulatory Guide 1.3 source term assumptions and l conservative PNPS-Unique parameters, the offsite thyroid and !
whole body doses for this limiting event are below 10CFR100 ;
limits. The extremely low probability of a fission product !
release, the minimal offsite radiological consequences of the TIP l containment isolation failure, and the prohibitive costs involved L in upgrading the position indicating circuits for the isolating TIP shear and ball valves support the Boston Edison decision not to upgrade the Category 3 equipment provided for this variable. j l
H. Radioactivity Concentration in Circulating Primary Coolant (Tvoe C. I Category 3) l The classification of this variable at PNPS as Category 3 deviates from the Regulatory Guide 1.97 recommendation of Category 1. j 4
Instrumentation-to monitor radioactivity concentration in circulating I primary coolant is designated as Category 3 because no planned ,
operator actions are identified and no operator actions are i anticipated based on this variable. The existing Category 3 l instrumentation provided by the post-accident sampling system (PASS) '
adequately measures radioactivity concentration in the coolant to indicate fuel cladding failure. In Reference 2, the NRC concluded that the alternative instrumentation provided by PASS was acceptable l to monitor this variable.
l I. ' Sunnression Chamber Sorav Flow (Tvoe D. Cateaorv 2) and Drvwell Soray Flow (Tvoe D. Cateaory 2)
Regulatory Guide 1.97 recommends dedicated, Category 2 flow indication be provided on both the suppression chamber and drywell spray lines. At PNPS, Category 2 flow indication is provided on the ,
residual heat removal (RHR) injection line which feeds the LPCI, i suppression chamber spray, drywell spray, and the suppression chamber I cooling lines. PNPS deviates from the Regulatory Guide 1.97 !
recommendation because dedicated flow indication is not provided on i each spray line. l Operation of the suppression chamber and drywell sprays at PNPS requires the operator to manually open valves which divert RHR system flow to the sprays. These valves are normally closed and each is ,
provided with Category 1 valve position indication in the control i room. The knowledge of valve positions, coupled with RHR flow indication, assures the operator that flow is being diverted as desired to the suppression chamber spray and the drywell spray. '
Additional verification that the suppression chamber and drywell sprays are operating as designed is indiri.ctly provided by the Category 1 instrumentation indicating primary containment pressure. >
During accident conditions, the emergency operating procedures direct the control room operators to verify primary containment pressure to confirm the operation of the containment spray subsystems. Primary Page 6 of 16
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l containment pressure indication tells the operator that the l containment spray system spargers are operating within 3 to 5 minutes af ter system initiation. The containment spray system causes the ,
primary containment pressure to decrease rapidly by approximately 16 ;
psig, according to the calculated pressure responses of the '
containment. l The RHR-flow and the injection valve position indications strictly provide the operator with the knowledge that there is flow and the ,
spray path is open. The primary contaiheent pressure indicators .l
~ assure the operator that the subsystems are working as intended, i Boston Edison concludes that the alternative instrumentation described above provides adequate indication of the suppression chamber and drywell spray flows, J. Main Steamline Isolation Valve (MSIV) Leakaae Control System Pressure
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(Tvoe D. Category 2.1 l Regulatory Guide 1.97 recommends pressure indication be provided for ;
the MSIV leakage control system. This Category 2. Type 0 variable is :
not applicable to PNPS because no designated leakage control system exists on the main steamline isolation valves (Reference 3).
K. Isolation Condenser System Shell-Side Mater Level (Tyne D. CatigDIY
- 2) and Valve Position (Tvne D. Category 2)
]
No isolation condenser system is provided in the Mark I containment design at PNPS: therefore, these variables are not applicable to PNPS.
L. Low Pressure Coolant Iniection (LPCI) System Flow (Tvne D. Category 2) l Regulatory Guide 1.9/ recommends dedicated, Category 2 flow l indication be provided for the LPCI system injection into the reactor '
vessel. At PNPS, Category 2 flow indication is provided on the '.
residual heat removal (RHR) injection lines. PNPS deviates from the Pegulatory Guide 1.97 recommendation because dedicated flow indication is not provided on the LPCI injection line.
Operation of the LPCI system is verified by RHR flow indication and LPCI injection valve position. The RHR flow indication has a range of 0 to 20,000 gpm. This is adequate to cover the required range of ,
O to 110% of the PNPS LPCI system design flow, which is 0 to 15,840 gpm. The injection valves on the LPCI system flow path are provided -
with Category 1 valve position indication in the control room. The knowledge of valve positions, coupled with RHR flow indications, assures the operator that flow is being sent, as desired, to the LPCI '
systeminjectionline.
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Additional verification that the LPCI system injection is operating :
as designed is indirectly provided by the Category 1 instrumentation ;
indicating reactor pressure vessel water level. During accident I conditions, the emergency operating procedures (EOPs) direct the ,
control room operators to verify reactor vessel water level to confirm the operation of the safety injection systems such as LPCI. t Boston Edison concludes that the alternative instrumentation '
described above provides adequate indication of the LPCI system flow.
M. Standby Liould Control System (SLCS) Flow (Tvoe D. Category 3)
Regulatory Guide 1.97 recommends that Category 2 flow indication be provided for the SLCS injection into the reactor vessel. At PNPS, proper operation of the SLCS is monitored by the Category 3 variables -
SLCS pump discharge header pressure and SLCS r.torage tank level, i The current design basis for the SLCS recognizes that the system has I
.. an importance to safety that is less than the importance to safety of L the reactor protection system and the engineered safeguards systems. ,
l Accordingly, the instrumentation provided to monitor the operation of '
the SLCS is considered to be Category 3 (Reference 1).
i The indication of SLCS pump discharge header pressure assures the o operator that the SLCS pumps are operating as designed. The :
1 instrumentation has a range of 0 to 2,000 psig, which sufficiently '
l encompasses the system design pressure of 1500 psig. All valves .
located between the SLCS storage tank and the reactor pressure vessel are normally locked open, with the exception of check valves and the ,
highly reliable squib valves. A reduction in the SLCS storage tank
. level indication assures the operator that the SLCS is actually pumping fluid into the reactor vessel. Boston Edison concludes that the alternative instrumentation described above provides adequate :
indication to monitor the operation of the SLCS (Reference 3), 1
- N. Standbv Liouid Control System (SLCS) Storage Tank Level (Tvoe D. .
.Categorv 3) i Regulatory Guide 1.97 recommends that Category 2 indication be provided for the SLCS storage tank level. At PNPS, Category 3 ,
instrumentation is provided to monitor this variable.
The current design basis for the SLCS recognizes that the system has i an importance to safety that is less than the importance to safety of the reactor protection system and the engineered safeguards systems.
Accordingly, the instrumentation provided to monitor the operation of the SLCS is considered to be Category 3 (Reference 1). 1 l-Boston Edison provided additional justification for this position to ,
, the NRC in Reference 3. Since then, the scale on the SLCS storage i
tank level has been replaced as a result of an enhancement identified by the Detailed Control Room Design Review (DCRDR) Project. The scale for this indication is now calibrated to read from 0 to 4,750 .
gallons- This new scale meets the intent of the Regulatory Guide l 1.97 recommended range of top to bottom of the tank.
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l O. Added Plant Variables (Tyne D. Cittgory 3) ]
Regulatory Guide 1.97 provides a recommended minimum set of plant variables that should be monitored during and following an accident. I At PNPS, this minimum set is supplemented by the following six plant variables. These variables provide important information to indicate L the operation of individual safety systems and other systems
- important to safety. This Category 3 instrumentation provides j indication in the control room for each plant variable. In the case '
of the additional drywell atmosphere temperature instrumentation, ,
indication is provided in the control room on the EPIC computer. l
- Bypass Valve Position
- Condenser Hotwell Level
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- Condenser Vacuum
- Condensate Flow e Recirculation Flow L
- Drywell Atmosphere Temperature P. Reactor Buildina or Secondary Containment Area Radiation (Tyne E. I Category 2)
Boston Edison's position is that this Regulatory Guide 1.97 e recommended variable is not required for the PNPS Mark I containment '
design.
l The exposure rate in the secondary containment will be largely l dependent on the radioactivity in the primary containment and the fluids flowing =through the emergency core cooling system (ECCS) l piping. Local radiation exposure rate monitors could only provide ambiguous indications because there are a large number of pipes in ;
widely scattered locations. The noble gas effluent monitors will '
provide a more appropriate means of detecting any radioactivity :
release. For these reasons, area radiation indication in the secondary containment would not provide the operator with useful
- information and is not required at PNPS (Reference 3), ,
Q. Ratta. tion Exposure Rate (Tyge_L Category 3)
, R U ulatory Guide 1.97 recommends a range of 10E-1 to 10E4 R/hr for instrumentation to monitor the radiation exposure rate in areas where L
access is required to service equipment importuc to safety. The :
installed instrumentation at PNPS has a range of 10E-5 to 10E-1 R/hr. i Boston Edison will use the existing area radiation monitors and supplement them, on an as-needed basis, with portable radiation monitoring equipment that exists onsite. Because the portable radiatior, monitoring equipment is fully capable of covering the range -
of radiation exposure comparable to the emergency condition allowable exposure limits (25 R for health, safety, and property protection and 75 R for life saving), this alternative to hardware modifications meets the Regulatory Guide 1.97 recommendation to monitor access areas required to service equipment important to safety (Reference 3).
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'R. Particulates and Halogens (Tvoa E. Categorv 3)
Regulatory Guide 1.97 recommends that a range of 10-3 pCi/cc to 10' gCi/cc be provided for instrumentation to monitor airborne radioactive materials (particulates and halogens) released from the plant. As described below, this is accomplished at PNPS through the r combined use of existing instrumentation (multi-channel analyzer systems and radiation monitor survey meters), procedures, and analytical tools in the form of nomograms. The combined ranges provided at PNPS to measure airborne radionuclide concentrations of L particu ttes and halogens re' eased from the plant is from 1 X 10- 4 pC1/cc to 3.5 X 104 pCi/cc, which encompasses the recommended range. )l In an accident condition, the identified release points at PNPS for particulates and halogens are the main stack, the reactor building l vent, and the turbine building. Releases from the main stack and reactor building vent are sampled through the use of a particulate ;
filter and a charcoal-based iodine collection chamber, installed i ahead of the routine effluent monitoring sample 'ines. For turbine j building releases under accident conditions, particulates and j halogens are sampled through the use of a portabl t air sample pump <
and filter (Reference 5). l l
Station procedures specify how samples of effluer,t particulates and i halogens will be collected and analyzed under accident conditions !
from the main stack, reactor building vent, and turbine building.
When the sample dose. rate is 125 mR/hr, the sample is measured in the onsite radiochemistry lab using a multi-channel analyzer. When the sample dose rate is > 25 mR/hr but 1 550 mR/hr, the sample may be 1 measured using the multi.-Gannel analyzer if it is first cut down to a section that has a dose rate 1 25 mR/hr. When the sample dose rate is > 550 mR/hr, the sample cannot be analyzed until it has decayed sufficiently (Reference 5).
The range of detection of the multi-channel analyzer is from 1 X 10-82 pCi/cc to 6.4 X 10-3 pCi/cc. The estimated upper limit of concentration can vary depending on the radionuclide species present and the elapsed time after reactor shutdown.
In addition to the multi-channel analyzer, station procedures require the "se of nomograms to estimate sample activity from the sample dose rate When the sample dose rate falls in the range 10-2 mR/hr to 104 '
R/hr the nomograms are capgble of est mating Iodine-131 inventory on the sample in the range 10-' pCi to 10 mci (Reference 5). The resultant range of Iodine-131 equivalent effluent plant release concentratigns estimated from the nomograms is from 3.5 X 10-0 pC1/cc to 3.5 X 10* pCi/cc.
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I, High range radiation survey instruments (Teletector or equivalent) are available to measure dose rates up to 103 R/hr. The radiation dose received by plant personnel in the collection, handling, transporting, and analyzing of effluent samples will not exceed the exposure limits of General Design Criterion 19 (Reference 5).
Boston Edison concludes that the overlapping ranges provided by the multi-channel analyzer and the nomograms sufficiently encompass the range recommended by Regulatory Guida 1.97.
S. Airborne Radiohalocens and Particulates (Tvoe E. Cateoorv 3)
Regulatory Guide 1.97 recommends that a range of 10-9 pCi/cc to 10-3 pCi/cc be provided for instrumentation to measure samples taken in the field for airborne radionuclide concentrations of particulates and halogens in the environs. As described below, this is accomplished at PNPS through the combined use of existing instrumentation (multi-channel analyzer in the onsite radiochemistry lab, SAM-2 sodium iodide detector with a dual-channel analyzer in the field, and radiation monitor survey meters both on and offsite),
procedures, and analytical tools in the form of nomograms. The combined range provided at PNPS to measure field samples for airborne radionuclideconcentratiggsofparticulatesanghalogensinthe environs is from 1 X 10- pCi/cc to 6.4 X 10- pCi/cc, which sufficiently encompasses the recommended range.
Field samples of airborne radionuclide concentrations of particulates and halogens in the environs surrounding PNPS can be measured using a SAM-2 detector in the field, a multi-channel analyzer in the onsite radiochemistry lab, or nomograms to estimate Iodine-131 equivalence until the samples can be brought to the onsite lab for analysis by the multi-channel analyzer.
The range of detection for the SAH-2 sodium iodide detectors in the field is from 8 X 10-9 pC1/cr +o 8 X 10-b pCi/cc. No quantitative measurement of. particulate filter paper sample activity is made in the field. However, estimates of Iodine-131 concentrations in the environs can be made in an expeditious manner using a ngmngram. When the sample count rate falls in the range of I cpm to 10' cpm, the nomogramiscapableofestigatingIodine-131inventoryonthesample in the range 10-0 pCi to 10' pCi. The resultant range of Iodine-131 equivalent (gncentration-in the environs estimated from the nomogram is from 10-14 pCi/cc to 10-4 pCi/cc.
The nomogram is used for quick Iodine-131 airborne concentration estimates by field teams, after which the field samples are brought back to the onsite radiochemistry lab for analysis using the multi-channel analyzer for accurate assessment. The ra detection of the multi-channel analyzer is from 1 X 10nge of to Ci/cc 6.4 X 10-3 Ci/cc. The multi-channel analyzer system is only used to Page 11 of 16
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l analyze samples whose contact gamme dose rate is 125 mR/hr, in 5 accordance with station procedures. The radiation dose received by i plant personnel in the collection, handling, transporting, and -
analyzing of field samples will not txceed the exposure limits of General Design Criterion 19.
Boston Edison concludes that the overlapping ranges provided by the ,
SAH-2 detector, the multi-channel analyzer, and the nomograms ;
sufficiently encompass the range recommended by Regulatory Guide 1.97.
T. Electrical Seoaration and Isolation ;
Regulatory Guide 1.97 requires that the redundant or diverse channels of Category *, equipment be electrically independent and physically separated from each other and from equipment not classified important ,
to safety up to, and including, any isolation device. Regulatory r Guide 1.97 references Regulatory Guide 1.75, " Physical Independence of Electric Systems" as the standard for this requirement. ,
L PNPS was designed and constructed to meet the proposed IEEE Standard
" Criteria for Nuclear Power Plant Protection Systems," dated March .
1968, which predates the issuance of Regulatory Guide 1.75.
The following separation criteria shall be used at PNPS, in '
accordance with Boston Edison Specification E-347, Section 5.4; Boston Edison Specification E-347A, Sections 5.2.3 and 5.2.4; and PNPS FSAR Section 8.9.3. These criteria are considered rainimum requirements and design guidelines for use in the absence of a -
confirming design review to support less stringent requirements.
l e Cable Trav Cable Spreading Room Area:
The_minimrm separation distance between redundant Class lE cable trays shall be 1 foot between-trays separated horizontally SM 3 feet between trays separcted vertically. Where plant arrangement ;
precludes maintrining the ininimum separation distanco between .
trays, barriers shall be provided betwekn redundant circutts.
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General Plant Areas: -
i The minimum separation distance between redundact Class IE ceble trays shall be 3 feet between trays 59parated borizontal)y *.nd 5 feet between trays separated vertically. Where plant ernngement precludes maintaining the minimum separation distance batuen trays, barriers shall be provided between redundant circuits. :
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e Enclosed Raceway-Cab 1(r Spreading Room and General Plant Areas:
The minimum separation distance between redundant Class 1E enclosed raceways shall be 1 inch. ,
i e Internal Wirina !
The minimum separation distance between control pancl internal wiring for redundant monitoring channels shall be 6 inches. ,
Where this separation cannot be maintained, a qualified barrier i shall be provided as described in IEEE Standard 384-1974.
BECo intends to use the guidance provided in Regulatory Guide 1.75, where applicable, with the following exceptions:
- The cable spreading room area contains instrumentation and control cables along with a 480V load center and a 480V motor ,
control center. The 480V cables are routed in conduit and cable j trays and separation shall be. maintained in accordance with the '
requirements of the cable spreading room area, as stated above.
e Raceway markings are located at various intervals to provide adequate raceway identification. Conduits are labeled where they pass through walls and floors, at the conduit destination and i origin points, and at other locations along the conduit. The interval between labels may exceed the 15-foot recommendation of I Regulatory Guide 1.75.
- Associated cables and racewhys are not uniquely identified.
< Unique identification is not required to ensure electrical l l separation of redundant systems. l l
l
- Electrical isolation shall be accomplished by use of coordinated I li Class 1E fuses or breakers, in accordance with the proposed IEEE l Standard, " Criteria for Nuclear Power Plant Protection Systems,"
dated March 1968 (Reference 1).
III. OPEN ITEMS II The open items indicated on Tables 1 and 2 require additional work to p verify compliance with Regulatory Guide 1.97, Revision 3. Open items o related to specific issues are discussed below.
L A. Eauioment Oualification Additional work is required on the environmental qualification of the p, instrumentation and associated equipment listed as open on Tables 1 l and 2. As discussed in Reference 5, the evaluation of ine '
l- post-accident environment and basis for qualification for the
. instrumentation and associated equipment monitoring effluent radioactivity and status of standby power will be completed and :
submitted to the NRC under separate cover.
Page 13 of 16 i
B. Seismic Oualification ;
As stated in Reference 1, Boston Edison deferred the review of the seismic qualification of accident monitoring instrumentation for l' Regulatory Guide 1.97 pending the resolution of Unresolved Safety Issue (USI) A-46. The generic letter stated equipment must either be !
qualified using seismic experience data in accordance with procedures !
developed by the Seismic Qualification Utility Group or by the -
analysis and testing methods of IEEE Standard 344-1975. ;
Subsequent to the generic letter, the Seismic Qualification Utility Group submitted a generic implementation procedure (GIP) for NRC '
approval. The GIP contains evaluation procedures and acceptance criteria for the use of seismic experience data in the resolution of USI A-46. The NRC issued its safety evaluation of the GIP in July. .
1988. The GIP applies to plants with construction permits issued prior to 1972 (i.e., plants not originally licensed to IEEE Standard 344-1975 at startup). PNPS is in the group of plants covered by the GIP.
L Related to this, IEEE Standard 344 was revised in 1987 to include ;
L provisions for the use of seismic experience data to qualify L electrical equipment. The standard applies to all plants regardless of age. The NRC endorsed this standard in the latest revision to Regulatory Guide 1.100 " Seismic Qualification of Electric Eouipment for Nuclear Power Plants."
In view of these developments Boston Edison has initiated a program to verify the seismic qualification of Regulatory Guide 1.97 Category 1 equipment and the Category 2 equipment of safety-related systems, Equipment purchased and installed to the requirements of IEEE Standard 344-1975 will be deemed acceptable as is, provided the qualificatirn documentation is readily available and auditable. For other equipment, the program will verify the seismic qualification per IEEE. Standard 344-1987 Section 9, Experience, and the Seismic Qualification Utility Group Generic Implementation Procedure, ;
Revision 1, dated November,1988. Upon completion of this qualification program, a summary of t5e results will be submitted to the NRC. !
C. Neutron Flux Monitorina As discussed in Section II.D. Boston Edison endorses the BHR Owners' Group's position that a fully qualified Class IE post-accident heutron monitoring system is not appropriate. Compliance for this item remains open pending completion of the NRC's review of the BWR Owners' Group position, i
L Page 14 of 16
~r-'vy" A us.-.m+e =+ _ . , , _ _ _ _ _ _ _ , , . _ _ _ . _ ___,,___,___,___.__,_,_.__________,__.__._,,_____,___,._,,_____m____,
F 1
- 3 ;
[ , . !
Eauinment Identification and Human Factors u , D. ;
L ;
In conjunction with the Boston Edison Detailed Control Room Design !
Review (DCRDR) Project, a human engineering review of Regulatory i Guide 1.97-related devices on the main control room panels will be j performed in accordance with NUREG-0700. The Regulatory Guide .
i r 1.97-related devices on the main control room panels will be marked or identified at such as part of the ongoing control room I enhancements activity. These activities are further described in l Boston Edison letters to the NRC, dated May 2, 1989 and July 24, ;
1989. The remaining Regulatory Guide 1.97-related devices outside i the main control room panels will also be reviewed and marked in a l similar manner. The DCRDR Project is included in the Doston Edison i Long Term Program. !
. 1 L- E. Channel Availability. Channel Redundanev. Quality Assurance. and ;
Testina )
Boston Edison is currently reviewing the compliance of the j post-accident monitoring instrumentation at PNPS with the Regulatory. 1 Guide 1.97 recommendations for these design criteria. The results of '
our review will be submitted to the NRC upon completion.
l i
1
)
I l
l I
1 l
1 l
l I
Page 15 of 16 I 1
REFERENCES
- 1. . Letter from H. D. Harrington (BECo) to D. B. Vassallo (NRC), dated November 1, 1984 (BECo 2.84.187), " Generic Letter 82-33: Regulatory Guide 1.97"
'i . Letter from J. A. Zwolinski (NRC) to W. D. Harrington (BECo), dated December 12, 1985 (BECo 1.85.372), " Generic Letter 82-33; Regulatory Guide 1.97 Request for Additional Information"
- 3. Letter from J. M. Lydon (BECo) to NRC, dated February 10, 1987 (BECo 2.87.021), " Additional Information Concerning Regulatory Guide 1.97"
- 4. Letter from D. G. Mcdonald (NRC) to R. G. Bird (BECo), dated January 24, 1989 (BECo 1.89.044), " Emergency Response Capability, Conformance to ;
Regulatory Guide 1.97, Revision 3, Request for Additional Information"
- 5. Letter from R. G. Bird (BECo) to NRC, dated April 11,1989 (BECo 2.89.053), " Response to Request for Additional Information, Emergency Response Capability, Regulatory Guide 1.97, Revision 3 (TAC 51119)"
F t
Page 16 of 16
4 9 TABLE 1 - PNPS REGULATORY GUIDE 1.97 COtfPLIANCE RIATRIX ~
Seismic Power Seperationf Ohsemel Charmel Equip Humen Varleble Deviations EO Oust so.orce isoletion 1 Range Redund Avell ID Factors Displey OA Testing TYPE A CAT 1 RANGE deviation. 'O A A A AWJ A A O O A A O See Sedion f* A ATMOSPHERE TEMPERATURE TYPE A CAT 1 RANGE deviation. A A A O AWJ A A O O A A O CONTAINMENT AND DRYWELL HYDROGEN CONCENTRATION TYPE A CAT 1 A A A O A A A O O A A O CONTAINMENT AND DRYWELL OXYGEN CONCENTRATION TYPE A CAT 1 A A A O A A A O O A A O PRIMARY CONTAINMENT PRESSURE - DRYWELL TYPE A CAT 1 A A A O A O A O O O A O PRIMARY
! CONTAINMENT i PRESSURE-SUPPRESSION POOL TYPE A CAT 1 A A A O A A A O O A A O RCS PRESSURE TYPE A CAT 1 RANGE deviation. A O A O AWJ A A O O A O O COOLANT LEVELIN REACTOR VESSEL TYPE A CAT 1 A A A O A A A O O A A O SUPPRESSION POOL WATER LEVEL TYPE A CAT 1 A A A O A A A O O A A O SUPPRESSION POOL WATER TEMPERATURE A - Acx:eptable, meets the RG1.97 design and qualification criteria. 9 O - Open, see descriptions in Sedion !!L See Section ILT for Boston Edison's position on mmpliance l
- AWJ -@ wth justdication. with electrical separation and isolation design criteria.
TABLE 1 l NR - Not required, no specific provision required in RG1.97 Table 1. Page 1 of 8
.. _ _ . __z_ _ _ _ . _ _ _ . . . _ _ _ _ . . _ _ _ . . _ _ _ _ _ _ _ . _ . - _ _ _ _ _ - . . _ . _ _ _ . . _ _ _ _ . . . _ _ _ _ _ _ _ _ _ _ _ . . _
1
. *7 TABLE 1 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX ~
Seismic Power Separation / Channel Channel Equip Humen
' Variable Deviations EO Oust Source Isoletion 1 Range Redund Avsil ID Factors Dispisy QA Testing TYPE B CAT 1- Waitingfor NRC O O O O A A A O O A O O reviewof BWROG NEUTRON FLUX - position. See Section APRM ll.D TVPE B CAT 1 Waitingfor NRC O O O O A A A O O A O O reviewof BWROG NEUTRON FLUX-SRM See Sedion TYPE B CAT 3 NR NR NR NR A NR NR NR O A A O CONTROL ROD POSITION
!?PE B CAT 3 NR NR NR NR A NR NR NR O A A O FCS SOLUBLE BORON WNCENTRATION TYPE B CAT 1 RANGE deviation. A O A O AWJ A A O O A O O COOLANT LEVEL IN REACTOR VESSEL TYPE B CAT None Not included in PNPS BWR CORE TEMPERATURE S @**
TYPEB CAT 1 A A A O A A A O O A A O RCS PRESSURE TYPE B CAT 1 A A A O A A A O O A A O DRYWELL PRESSURE TYPE B CAT 3 aded variable NR NR NR NR A NR NR NR O A A O from 1 to Cat 3.
DRYVELL SUMP See Sedion ILF
! LEVEL TYPEB CAT 1 A A A O A A A O O A A O PRIMARY CONTAINMENT PRESSURE -DRYWELL TYPE B -CAT 1 A A A O A O A O O A A O PRIMARY CONTAINMENT PRESSURE-SUPPRESSION POOL TABLE 1 Page 2 of 8 c _ _ _ _.
.~ ,
.* -v-TABLE 1 - PNPS REGULATORY GUIDE 1.97 COtSPLIANCE MATRIX Seismic Power Soperationf Channel Channel Equip Humen .
Variable Devletions EQ Oust Se srce Isolation 1 Range Redund Avelt
. ID Factors Displev . oA Tm- ,
TVPE B CAT 1 See Ta.6le 2 for PCIV PRIMARY $"Section$
CONTAINMENT deviations.
ISOLATON VALVE POSITON TYPE C CAT 3 Downgraded vanable NR NR NR NR A NR NR NR O A A O
. from Gat 1 to Cat 3.
RADOACTIVITY See Section II.H CONCENTRATON IN CIRCULATING PRIMARY COOLANT TYPE C CAT 3 NR NR NR NR A NR NR NR O A A O ANALYSIS OF PRIMARY COOLANT TYPE C CAT None Not included in FNPS RG1.97 program. See BWR CORE Section ILE TEMPERATURE ,_
TYPE C CAT 1 A A A O A A A O O A A O RCS PRESSURE TYPE C CAT 3 NR NR NR NR A NR NR NR O A A O PRIMARY CONTAINMENTAREA RADIATON j TYPE C CAT 3 Ovm ye vanable NR NR NR NR A NR NR NR O A A O i
from Cat 1 to Cat 3.
DRYWELL DRAIN See Section ILF SUMPS LEVEL TYPE C CAT 1 A A A O A A A O O A A O SUITRESSON POOL WATER LEVEL TYPE C CAT 1 A A A O A A A O O A A O DRYWELL PRESSURE TYPE C CAT 1 A A A O A A A O O A A O RCS PRESSURE TYPE C CAT 1 A A A O A A A -O O A A O PRIMARY ,
CONTAINMENT PRESSURE -DRYWELL TABLE 1 Page 3 of 8
,; c-
+
TABLE 1 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX Seismic Power Seperation/ Channel Channel Equip Humon ,
Variable Deviations EO Quel Source isoletion 1 Range Redund Avail ID Factors Display OA Testing TYPE C CAT 1 A A- A O A O A O O A A O PRIMARY CONTAINMENT PRESSURE-SUPPRESSION POOL TYPE C CATt RA i:deviatiort A A A O AWJ A A O C- A A O CONTAINMENT AND DRYWELL HYDROGEN CONCENTRATON TYPE C CAT 1 A A A O A A A O O A A O CONTAINMENT AND DRYWELL OXYGEN CONCENTRATION TYPE C CAT 3 NR NR NR NR A NR NR NR O A A O CONTAINMENT EFFLUENT RADOACTIVITY -
NOBLE GASES TYPE C CAT 2 O NR A NR A NR O O O A O O EFFLUENT l
RADIOACTIVITY -
NOBLE GASES TYPE D CAT 3 NR NR NR NR A NR NR NR O A A O MAIN FEEDWATER FLOW TYPE D CAT 3 NR NR NR NR A NR NR NR O A A O CONDENSATE STORAGE TANK LEVEL TYPE D CAT 2 RHR system flow and RHR to suppression SUPPRESSON poolspray valve CHAMBER SPRAY position used. See FLOW Section II.I TYPE D CAT 2 A A A NR A NR A NR O A A O DRYWELL PRESSURE TYPE D CAT 2 A A A NR A NR A NR O A A O SUPPRESSON POOL
, WATER LEVEL TABLE 1 Page 4 of 8
. ._ o.
e
.. e'
~
TABLE 1 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE RBATRIX Solomic Power Separation / Cliennel Caennel Equip Human .
Verieble Devletions EO Quel Source isoletion 1 Range Redund Aveil ID Factors Display OA Testing TYPE D CAT 2 A A A NR A NR A NR O A A O SUPPRESSION POOL WATER TEMPERATURE TYPE D CAT 2 RANGE deviation. O A A- .NR AWJ NR A NR O A A- O See Section ll.A ATMOSPHERE TEMPERATURE TVPE D CAT 2 RHR system flow and RHR to drywel spray DRYWELL SPRAY valve position used. .
FLOW See Section IU Not included in the
~
TYPE D CAT 2 PNPS Mark idesign.
MSIVS LEAKAGE See Section II.J CONTROL SYSTEM PRESSURE TYPE D CAT 2 O O A NR A NR O NR O A O O PRIMARY SYSTEM SAFETY RELIEF VALVE POSITIONS TYPE D CAT 2 Not included in.the
- PNPS Mark I design.
ISOLATION See Sedion II.K CONDENSER SYSTEM SHELL-S1DE WATER LEVEL TYPE D CAT 2 Not included in the PNPS Mark Idesign.
1 ISOLATION See Sedion II.K CONDENSER SYSTEM VALVE POSITION TYPE D CAT 2 O O A NR A NR A NR O A O O RCIC FLOW TVPE D CAT 2 A O A NR A NR A NR O A O O HPCI FLOW TYPE D CAT 2 O O A NR A NR O NR O A O O CORE SPRAY SYSTEM FLOW TABLE 1 i Page 5 of 8
~
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~
TABLE 1 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX Seismic Power Separationf Charmel . Channel Equip Humon ,
Variable Deviet6cns EO Oust Source tooletion 1 Range Redund Avoit ID Factors Displey OA Testing TYPE D CAT 2 RHR system flow and LPCIinjection vane LPCISYSTEM FLOW position used.See i Section ILL -
TYPE D CAT 3 aded v6nable NR NR NR NR A NR NR NR O A O O from 2 to Cat 3.
SLCS FLOW SLCS pump dischargepressure is used. See Sect'on itM TYPE D CAT 3 aded variabb NR NR NR NR A NR NR NR O A O O from 2 to Cat 3.
SLCS STORAGE TANK See Section ILN LEVEL TYPE D CAT 2 O O A NR A NR O NR O A O O RHRSYSTEM FLOW TYPE D CAT 2 O O A NR A NR O NR O A O O RHR HEAT EXCHANGER OUTLET TEMPERATURE TYPE D CAT 2 O O A NR A NR O NR O A O O
- COOLING WATER TEMPERATURE TO ESF SYSTEM COMPONENTS l
TYPE D CAT 2 O O A NR A NR O NR O A O O COOLING WATER FLOW TO ESF SYSTEM
- COMPONENTS I TYPE D CAT 3 NR NR NR NR A NR NR NR O A A O l HIGH RADIOACTIVITY l LIOUID TANK LEVEL j TYPEO CAT 2 O NR A NR A NR O NR O A O O EMERGENCY VENTILATION DAMPER POSITION TABLE 1 Page 6 of 8 l
w- - --
,. +
e.
TA98_E 1 - PNPS REGULATORY GUIDE 1.37 COGAPLIANCE RfATRE .
Seismic Power Seperothmf Chennel Cherw=el Equip h .
ID Factors Displey OA Testing Deviations EO Ouel Sovree tooletion 1 L;;;; Redund Aveil Var' e tt A O O O NR A NR O NR O TYPE D CAT 2 O .A STATUS OF STANDBY POWER AND OTHER SOURCES OF ENERGY IMPORTANT TO SAFETY A A O NR NR NR NR' NR NR NR O TYPE D CAT 3 Added plant specife NR variable for rnore BYPASS VALVE ,nformgiort. See POSITION Section ILO NR NR NR NR O A A O TYPE D CAT 3 Added plant specific NR NR NR NR variable for more CONDENSER ,nformgion.See HOTWELLLEVEL Section gl.O NR NR NR NR O A A O TYPE D CAT 3 Added plant specific NR NR NR NR vanable for more CONDENSER VACUUM information.See SedLi ILO O NR NR NR NR O A A Added plant specife NR NR NR NR TYPE D CAT 3 variahis for more CONDENSATE FLOW information.See Section ILO NR NR NR NR O A A O C AT 3 Added plant specife NR NR NR NR TYPE D variable for more RECIRCULATION informgion. See FLOW Section ILO A O NR NR NR NR NR O A CAT 3 Added plant specific NR NR NR
$ TYPE D
~
variable for more DRYWELL information. See ATMOSPHERE Section ILO TEMPERATURE A O O A A A O O A TYPE E CAT 1 A A A PRIMARY CONTAINMENT AREA RADIATION - HG1 RANGE TYPE E CAT 2 Not reqtrired at PNPS, REACTOR BUILDING OR SECONDARY CONTAINMENTAREA RADIATION TABLE 1 Page 7 of 8
.c '
c 4-
, TABLE 1 - PNPS REGULATORY GUIDE 1.97 CORAPLIANCE AAATRE seismic power seperstient chorm.: Charmet Equip Human .
Verlable Deviettons EO Quel Source Isoletion 1 Range Redund Avail ID Factors Displey OA' Testing TYPE E CAT 3 RANGE devehort. NR NR NR NR AWJ- NR NR NR O A A O RADIATION EXPOSURE RATE TYPE E CAT 2 O NR A NR A NR O NR O O O O NOBLE GASES AND VENT FLOW RATES (COMMON PLANT VENT)
TVPE E CAT 3 See Sedion ILR for NR NR NR NR- A NR NR NR O A A O PARTICULATES AND onr .
HALOGENS TYPE E CAT 3 See Section ILS for NR NR NR NR A NR NR NR O A A O additionalinformation AIRBORNE on range.
RADIOHALOGENS AND PARTICULATES TYPE E CAT 3 NR NR NR NR A NR NR NR O A A O PLANT ENVIRONS RADIATION (PORTABLE)
TYPE E CAT 3 NR NR NR NR A NR NR NR O A A O PLANT ENVIRONS RADIOACTIVITY (PORTABLE)
TYPE E CAT 3 NR NR NR NR A NR NR NR O A A O METEOROLOGY TYPE E CAT 3 NR NR NR NR A NR NR NR O A A O PRIMARY COOLANT AND SUMP TYPE E CAT 3 NR NR NR NR A NR NR NR O A A O CONTA!NMENT AIR 4
TABLE 1 Page 8 of 8
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^
TABLE 2 - PNPS REGULATORY GU_IDE 1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVES Seismic Power Separationt Channel Channel Equip Humen
- Valves Deviation EQ Qual Source Isolation 1 Range Redund Avail ID Factors Display QA Testing AO203-1 A A O A O A A O O O A A O AO203-2A MSIV UNE *A-AO203-1B A O A O A A O O O A A O AO203-2B MSIV UNE 8-l AO203-1C A O A O A A O O O A A O AO203-2C MSIVUNE C-A0203-1D A O A O A A O O O A A O AO203-2D MSIVUNE D-MO220-1 O O A O A A O O O A A O MO220-2 MAIN STEAM DRAIN M01201-80 REDUNDANCY O O A O A AWJ O O O A A O (Check valve 6-58A) deviation. See S di II.G.1 RNN MO1301-49 REDUNDANCY O O A O A AWJ O O O A O O (Check valve 1301-50) deviaten See
- ~
RC C PUMP DISCHARGE MO23018 REDUNDANCY O O A O A AWJ O O O A O O (Check valve 2301-7) deviaton. See
! HPCI PUMP DISCHARGE MO1001-47 O O A O A A O O O A O O MO1001-50 RHR S.D COOUNG MO1201-2 O O A O A A O O O A A O M01201-5 RWCU SUCTICN e SV5065-31B O O A O A A O O O A O O SV5065-358 H2 O2 ANA!_YZER SUPPLY See Section II.T for Boston Edison's positen on compLees TABLE 2 O- see d a AWJ - Acceptat4e with justification. &Walwah MWon &sWW %1M NR - Not required, no specife provision required in OG1.97 Table 1.
3 -+ a o
_ TABLE 2 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVES Solomic Power Seperation/ Channel Channel Equip Humen Velves ' Deviation EQ - Oesi Source isoletion 1 Range Redund Avail ID Factors Display OA TestiEg MO1400-24A O O A O A A O O O A A O MO1400-25A CORE SPRAY TO REACTOR MO1400-24B O O A O A A O O O A O O MO1400-25B CORE SPRAY TO REACTOR MO1001-60 Dsarmed vahres.
MO1001-63 Net part d RG1.97 RHR HEAD SPRAY @cn I'l 2a AO7017A O O O O A A O O O A O O A070178 RMf COLLECTION AND DM FLOOR SUMP AO7011A O O O O A A O O O A O O AO70118 R/W COLLECTION AND j D/W FLOOR SUMP MO4002 REDUNDANCY O O A O A AWJ O O O A O O deviation. See RBCCW RETURN Sedion II.G.1 AO5043A O O A O A A O O O A O O j AO50438 DRYWELL2 EXHAUST BYPASS AO5044A O O A O A A O O O A O O
- AO50448 j DRYWELL PURGE l EXHAUST SV5081A O O A O A A O O O A O O SV5081B POST ACCIDENT PURGE AND VENT SV5082A O O A O A A O O O A O O SV5082B POST ACCIDENT PURGE AND VENT TABLE 2 Page 2 d 8
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66 AR 22 HT B U G CT CT lv N L 37 AP 12 A s3A L e 35 EE A8 55 R 55 U ABAECN66 A8AE CN3v Aa LL 33 NU 99 A N 33 L 77 33 E 11 22 I T v 00 WG 00 P 00 TV 00 66 OR 77,L A 77, H 00 W 30 33 88 55 SD 5588SDTV 5e3k 0c AE RY MV 55 55 0, AS 66 00 S 55 2 U AC B 2 S 00 AC S 33 55 9T E 77 33 P l
e 55 YR 55 I VV D V OO RU OO W VV ON VV ON Oh AA DP AA D SS PA SS PA A( ND C OR VV 2A VV SS HP CC CR 77 T 1
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TABLE 2 - PNPS REGULATORY GUIDE '1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVE!I Seismic . Power Separation / .
Channel Channel Equip Hurnen ,.
wi Valves ' Deviation EQ Qual ' Source isolation 1 Range Redund Avsil ID Factors . Dispisy OA Testing -
- SV302-121 Not part d RG1.97 -
.SV302-122 . program. See CRD WITHDRAW
' MO1001-23A O O A .O A A O O O A A' O -
MO1001-26A RHR TO DRYWELL SPRAY M01001-23B O O A O A A O O O A ~O O MO1001-26B RHR TO DRYWELL SPRAY SV5065-63 O O A O. A A O O O A O O SV5065-64 PAS RX SAMPLE SV5065-85 O O A O A A O O O A O O SV5065-86 PAS RX SAMPLE AO220-44 O O A O A A O .O O A A O AO220-45 REACTOR SAMPLE LINE SV5065-24A O O A O A A O O O A O O SV5065-26A H2!02 AND PASS GAS RETURN SV5065-138 O O A O A A O O O A O O.
SV5065-208 H2.02 ANALYZER SUPPLY MO1001-28A O O A O A A O O O A A O MO1001-29A LPCIINJECTION MO1001-28B O A O A A O O -
A A O MO1001-298 LPCIINJECTION MO2301-4 O O A O A A O O O A- O O MO2301-5
- HPCI TURBINE STEAM j SUPPLY
. TABLE 2 -
Page 4 d 8
-. . _ _ _ _ _ . . _ , _ _ _ _ . _ _ _ _ . . _ ~ _ _ - _ _ . ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ -. .
.. . _,=__ - ._ - _ _ - _ _ _ _ = ___ __-. ...
._ ~ mm
^
^
- ,. Wi3 6
TABLE 2 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVES Seismic Power . Separation / . Channel. Channel Equip ~ Human
' Velves Deviation Ef4 - Quel Source Isolation 1 Ronce RedJnd -- Avsil ID ' Factors DIspisy OA : TestN
'A MO1301-1G . O- - O .-
~
O A .A O O 'O A. O~ O -.
~ M01301-17 RCIC STEAM TO -
TURBINE SV5065-14A O O A O A A O .O O A -O O SVE065-21 A .
H202 ANALYZER SUPPLY AO50338 O O A O A A O .O O A O. O.
AO5036A DRYWELUTORUS PURGE AO5036A O O A O A A O O O A O O AO50368 TORUS PURGE INLET SV5087A O O A O A A O O O A O- O SV5087B POST ACCIDENT PURGE AND VENT SV5088A O O 'A O -A A O O O A O .O SV50888 POST ACCIDENT PURGE AND VENT AO5033C REDUNDANCY O O A O A AWJ O O O A O O.
(Check valve 9-CK-341) deviation. See
"~ ^
NORMAL N2 MAKEUP TO 4
_ SUPPRESSION POOL MO1001-36A Terminate below MO1001-36B suppression pool Not part of RG1.97 RHR TEST RETURNS program.See
! Section ll_G.2.c MO1001-18A Terminate below M01001-188 suppression pool Not part of RG1.97
, RHR MINIMUM FLOW program.See Section li.C2.c i MO1001-34A O O A O A A O O O A A O MO1001-37A RHR TO SUPPRESS!ON POOLSPRAY TABLE 2 Page 5 of 8
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- w
~
TABLE 2 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE' MATRIX ' FOR PRIMARY CONT *2"*"'NT ISOLATIOtJ VALVES ~
Seismic ' Power . Separation / . Channel ~ Channel Equip . Humsn . .
Valves -- Deviation EQ Qual Source ' iso:stion 1 Range Redund Avat! ID - Factors ' Display OA ' Testi6g -
MO1001 '4B O O A- O A- A- ^O O- 'O A O. O
. MO1001-378 RHR TO SUPPRESSION ,
POOLSPRAY MO2301-33 O ~ O -. A O' -A A O O O A O O MO2301-34 HPCITURBINE EX VAC BRKR CV90E8* O O O O O O O O O O G~ O CV9M 38
. HPCIGLAND SEAL CONDENSER MO1301-25 Terminate below suppression pool RCIC PUMP SUCTION Not part of RG1.97 FROM TORUS program. See Section ILG.2.c
' MO2301-36 Terminate below suppression pool HPCI PUMP SUCTION Not part of RG1.97 FROM TORUS program. See Sedion ILG.2.c MO1001-7A,7B,7C,7D Terminate below -
suppression pool RHR PUMP SUCTION Not part of RG1.97 program. See
- Section ILG.2.c AO5040A REDUNDANCY O O' A O A AWJ O O O A O O (Check valve X-212A) deviation. See TORUS VACUUM BREAKERS ISOLATION AO5041A O O A O A A O O O A O O AO5041B TORUS EXHAUST BYPASS AO5042A O O .A O A A O O O A .. A O AO5042B TORUS MAIN EXHAUST AO5025 O O A O A A O O O A A 'O AO50428 DIRECTTORUS VENT ISOLATION _
' TABLE 2 Page 6 of 8
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, .. - g TABLE 2 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVES ~. ,
~~
Seismic Power Separation / - Channel . Channel ' Equip Human-Valves ' Deviation EO ' Qual - Source Isolation 1 Range Redund Avall ID Factors Display : OA Test 4
_ SV5083A- - O O- A- ;O A- A. O: .O O A- .O O SV50838 POST ACCIDENT PURGE
- AND VENT SV5084A : O- O A 'O- .A A O .O: '
O A O O SV5084B
-POS1 ACCIDENTPURGE
' AND VENT AO5040B REDUNDANCY. O. O A O A 'AWJ O O O A- O 0 (Check valve X-2128) - Joviation. See TORUS VACUUM BREAKERS 13OLATON SV5065-158 O O A O: A_ A O O O A O O SV5065-228 H2O2 ANALYZER SUPPLY CV5046 REDUNDANCY O O O O O AWJ O O O O 'O O (Check valve 31-CK-434) deviation. See AIR TO DW TO TORUS VACUUM BREAKERS SV5065-77 O O A O A A O O O- A O O SV5065-78 PAS LIQUlO RETURN SV5065-71 O O ~A O A A O O O A O O SV5065-72 PAS LIQUID RETURN SV5065-11 A ' O O A O A A O O O A O O.
SV5065-18A H2O2 ANALYZER SUPPLY SV5065-25B O O A O A A O O O A O 'O i SV5065-27B H202 ANALYZER SUPPLY MO1400-3A Terminate below M01400-3B suppression pool Not part of RG1.97
- CORE SPRAY SUCTON program. See Section ll.G 2.c TABLE 2 Page 7 of 8
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. -(af TABLE 2 - PNPS REGULATORY GUIDE 1.97 COMPLIANCE MATRIX FOR PRIMARY CONTAINMENT ISOLATION VALVE 5' '
Seisudd Power . Separation / _
Range Redund Channel .Avait Channel' ID l Equip' Human; Factors Display OA ' TestMg
[" -
Valves ' Deviation EO' Oual ' Source - Isolation ~1
- MO1001-21 . Not part of RG1.97 -
MO1001-32 : . program. See
_ RHR DISCHARGE TO
- RADWASTE t
TABLE 2 Page 8 of 8 m + , .- ;,,,y- ,.,..-.3 4 .. . . _ , m _m-4. , - _ - . . _ , - - - .. . .m._.__ _ _ _.....; ,_m.___ _
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