ML18052B520
| ML18052B520 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/31/1987 |
| From: | Johnson B CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NUDOCS 8804050311 | |
| Download: ML18052B520 (25) | |
Text
OC0388-0078-NL02 8804050311 871231 PDR ADOCK 05000255 R
DCD e.* e ATTACHMENT Consumers Power Company Palisades Plant Docket 50-255 PALISADES PLANT ANNUAL REPORT TO THE NRC FOR 50.59 REVIEWS March 31, 1988 25 Pages*
MI0188-265A-TC07 FACILIITY CHANGES, SPECIFICATION CHANGES, TESTS AND EXPERIMENTS DECLARED OPERAB~E IN 1987 CLOSED IN 1987
2 SC-87-121 This specification change covers the removal of the Exciter Voltage Regulator Volts/Hertz Card from the circuit.
The Hi Volts/Hertz alarm contact is still in operation without the card.
The card was causing machine instability as the machine voltage approached 23.1 KV.
Removal of the output from the card eliminates automatic limiter, while the Hi Volts/Hertz alarm is still operable and operates as m~chine rating is approached.
SAFETY EVALUATION
SUMMARY
FSAR Section 8.4.3.1 states "During coastdown* excitation control equipment functions to.main.tain a constant ratio of volts to frequency to maintain the excitation losses of transformers and rotating machinery within the thermal rating of equipment."
The Volt/Hertz limiter function has been removed.
There still exists a 5.9-81 relay that* provides a 110% limit on Volts/Hertz to prevent magnetic saturation in generator loads.
Within.the 110% over-volt/hertz, the Station Power Transformer 1-1 main back and PCP motors are capable of withstanding this condition during a coastdown evolution.
Coast-down will still be available and function as required. '
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
Over-excitation protective devices will allow at least 30 seconds of coastdown
- if it were.required.
The probability of.an accident or malfunction of a different type than any evaluated previously in. the FSAR is not crea'ted.
The margin of safety as defined in the basis for any technical specification is not reduced because PCP coastdown and voltage regulator are not referenced in the technical specification in relation to Volts/Hertz.
SC-83-138 This specification change replaced the fuel elevator right.angle drives.
This replacement was made due to the fact that the existing drives were designed by the Vendor for 78 inch pounds.
They should be designed for 300 inch pounds torque.
SAFETY EVALUATION
SUMMARY
The existing fuel elevator right angle drive units were replaced with heavy duty right angle drive units, designed jor increased torque to prevent failure of elevator drive train.
The probability of occurrence'-or the consequences of an accident or--malfunction of equipment important to safety previously evaluated in the FSAR is not increased by this modification.
The replacement equipment is identical in function to the original equipment.
The possibility of an accident or malfunction of a different type than previously evaluated in *the FSAR is not created.
MI0188-265A-TC07
SC-86-323 The modifications covered by this specification change included:
- 1. Modification to reduce the maximum flow to CCW Heat Exchangers E-54A and E-54B to prevent tube vibration.
- 2.
Modification to reduce the flows to the ESF pumps, to be consistent with the manufacturers requirements.
- 3. Modification to reduce the flow to the spent fuel cooler, reducing the total system flow.
- 4. Modification to the shutdown cooling heat exchangers flow, to r~duce.
total system flow.
SAFETY EVALUATION
SUMMARY
3 The CCW Heat Exchanger is capable of supplying 4500 gpm throughout without any apparent vibration problem.
This is confirmed by YUBA letter dated 11/10/86.
At 4500 gpm, the heat exchanger is expected to be operating near.an acceptable design limit.
This is supported by the operating history of the CCW Heat Ex-changer.
The heat exchanger experienced a failure of a number of tubes iI1 1970. Tubes were plugged between 1970 and 1979.
Not all the tubes plugged experienced leakage.
In 1975 a number of tubes were plugged in the inlet area of the heat exchanger to prevent future leaks.
The tupes had a rod installed and plugged.
The rod prevents further damage and prevents the inlet tubes*
from exciting others.
Since 197.9, there has been no leakage other than at previously plugged tubes*.
In 1981, the tubes in E-54B were eddy current tested with no defects.
The flows through the heat exchangers exceeded 4500*
gpm a number of times with no failures.
Based on the above, the probability of occurrence or the consequences of an accident or malfunction of equipment import~nt to safety previously evaluated. in the FS~ is not increased~
To limit flow through the CCW Heat Exchangers, flow has been reduced to the following components:.
- 1.
A reduction from 8000 to 5000 gpm to the shutdown coolers.
- 2.
A reduction in the minimum permitted flow to the HPSI pumps from 35,gpm/pump to 14.5 gpm/pump.
- 3.
A reduction in flow to the spent fuel pool cooler. __ The system will be throttled t_o_approximately 350 gpm.
The change in flow to the shutdown coolers has been analyzed by Combustion Engineering and found acceptable.
The change in flows to the HPSI pumps has been evaluated by Event Report, E-PAL-86-083 and Deviation Report D-PAL-86-192.
The reduction in flow to the spent fuel pool has been analyzed in calculation EA-E-PAL-86-083 and shown acceptable.
MI0188-265A-TC07
4 E-S4 Component Cooling Water Heat Exchanger flow is being reduced by reducing the required flow to the shutdown heat exchangers.
This reduction in flow has been analyzed and found acceptable.
The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.
The flow to shutdown heat exchangers has been reduced from 8000 to SOOO gpm.
However, analysis has demonstrated that, even with this reduced flow:
- 1.
Peak containment pressure does not exceed the peak design pressure of SS psig and pressure at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is l.ess than half.
These are the basis of the Standard Review Plan.
- 2.
Cooldown to 210°F in the required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is still achievable.
- 3.
Plant equipment can still be demonstrated to meet EEQ requirements.
In the Accident Analysis Review, the only other CCW requiremen~ is.Steam Generator Tube Rupture foll<;>wup to hold decay heat through shutdown cooHng at three and one-half hours.
This can still be met.
Therefore, defining the margin of.safetyas the difference between physical failure threshold and the Standard Review Plan criteria, the margin of safety is not reduced as defined in the basis for any technical specifications.
FC-571 This facility change covered the repair of SIRW tank cellular slab to bring the stress levels within the allowable limits stipulated in-the ACI 318-63 concrete code and the Palisades FSAR, thereby resolving the stress problems in the structure.
The repair includes filling all eight openings in the north/south beams with reinforced aggregate grout mix, sealing cracks and concrete covering exposed reinforcing steel.
The repair includes all work associated with filling openings and rerouting piping and conduit.
SAFETY.EVALUATION
SUMMARY
The repair of the SIRW tank cellular slab brings the stress levels within the allowable limits stipulated in the ACI.318-63 concrete code, and the Palisades FSAR, thereby resolving stress problems in the structure.
Repair includes filling eigllt beam openings with reinforced aggregate grout.mix, sealing cracks and concret'f!.. covering exposed reinforcing steel.
Repair includes all associated work such as~rerouting piping and conduit that would be inaccessible once cell openings are filled.* AIR A-P~-84-047 ~valuated the rerouting of piping in the slab for iodine removal makeup sodium hydroxide tank (T-103) and the iodine removal hydrazine tank (T-102).
The evaluation determined that the safety analysis assumed hydrazine concentration of at least 50 ppm in the containment sump following a LOCA would. meet with the proposed piping modification on T-102, and that there is also no impact on the limits imposed by Technical Specifica-tion 3.19.1.
The AIR also determined that the piping change on tank T-103 had no effect on the ability to supply NaOH to the sump.
Therefore, the possibility of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
MI0188-265A-TC07
5 Water from the SIRW Tank T-58, will be stored during a portion of the refueling outage in the clean waste receiver tanks, and the primary makeup water tank.
A radiological assessment of storing the contaminated water in T-90 was per-formed which approved the transfer and storage of water with recommendations given to control transfer and storage. Therefore, the possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created, and the margin of safety as defined in the basis for any technical specification is not reduced.
FC-601 This modification installed a water flow switch on the two *and one-half inch fire protection system line servicing the'electrical equipment room sprinkler system.
Under this facility change, this modification was "temporary," and the permanent installation will be done under FC-755.
Under FC-601, the flow switch was temporarily wired in parallel with another existing water flow switch due to the lack of spare annunciator windows on C-47.
FC-755 modification covers the installation of a replacement C-47 panel.
The balance of the electrical installations required by FC-601 have been completed.
SAFETY EVALUATION
SUMMARY
The probability of an undetected fire in the electrical equipment room is de-creased by this modification due to the addition of water flow switch actuated alarm for this area.
The method of mounting the actual detector will be the same as for presently mounted det.ectors.
Because of the light weight "and physi-cal configuration of the detector, no unacceptable stresses will be induced into the fire protection line due to eith.er static or seismic conditions.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The addition of another fire detection instrument provides improved fire detection capability and thus increases the margin of safety as defined in the basis for any technical specifications.
FC-683 This FC was closed out in 1986, and was not previously reported.
Although it is not reportable under 10CFRS0.59, it is being incorporated into the 1988 Annual Report because it involved an FSAR Change.
This modification covered the removal oi the pressurizer heater load centers from safety injection system (SIS) trip, and added proper annunciation of block relays to C13L-and Cl3R annunciator panels.
SAFETY EVALUATION
SUMMARY
This project removed the pressurizer heater load centers B14 and B15 from the SIS trip.
This change gave the operators another means to respond to various accident scenarios.
MI0188-265A-TC07
.. ------~---
Originally the pressurizer heaters were installed on SIS trip to protect the electrical penetration from a fault (suhmerge electrical equipment).
6 In 1983 the System Protection and Laboratory Services group performed a complete
- review of the electrical penetrations for adequacy of protection for overloads and short circuits.
This study was reviewed and the results have been documented.
From this review, we felt that by removing the pressurizer heaters from SIS trip we were able to know that the two electrical penetrations will still maintain their ~ntegrity for fault at the load center.
With the removal of this circuit (SIS trip) it is still possible to annunciate the blocked relays because of required technical specification testing.
This was done.with minor circuit modifications.
The above modif i~ations have not increased the probability of occurrence or consequences of an accident to equ'ipment.important to safety previously evaluated in the FSAR.
The removal of the pressurizer heaters from SIS did not create the possibility of an accident of a different type than any previously evaluated in the FSAR because the FSAR does not address the configt.tration of the heaters as being on Safety Injection System contacts for load shedding.
The technical specifications do not specifically identify any _of the equipment or circuits that were removed or modified under this project. *Therefore, the margin of safety as defined in the basis for technical specification was not reduced.
SPECIAL TEST T-223 The purpose of this test is to balance the CCW System in DBA condition such that the shutdown heat exchangers have a minimum flow of 5000 gpm, and to balance the*
CCW System in normal shutdown condition such that the flow through the CCW heat exchangers is less.than 4000 gpm each.
NOTE: Special Test T-223 was originally performed in 1986.
In 1987, only a s'mall portion of the test was performed.
SAFETY EVALUATION
SUMMARY
This test w,111 be performed during cold shutdown.
PCS temperature will be maintained between 70-150°F.
Spent fuel poet temperatures will be maintained below 130°F *. The probability of an accident is not increased because the equipment in the-component cooling system will be operated within their design limits.
The consequences of an accident are reduced because the plant is in cold shutdown and adequate CCW flow to safety related equipment is verified.
'This test will verify that a minimum of 5000 gpm is flowing to the shutdown heat exchangers, 61 gpm is flowing to engineering safeguards pumps, and 41 gpm is flowing to the charging pu~ps in a post~DBA condition.
MI0188-265A-TC07
7
-*~-...:-
The test will also verify that. a maximum of 4000 gpm is flowing through each ccµ heat exchanger with a minimum of 5000 gpm to the shutdown heat exchangers, 61 gpm is flowing to the engineering safeguard pumps, 41 gpm to the charging pumps and a minimum of 300 gpm to the spent fuel pool heat exchanger for normal plant shut-down.
The other CCW loads will be monitored to show adequate CCW flow.
Engineering safeguard heat exchangers, engineering safeguard pumps and spent fuel pool temperatures will be set lower than FSAR Table 9-7 requirements.
These are justified by Combustion Engineering analysis, revised vendor infor~
mation and fuel pool temperature rise without cooling data.
Temporary gauges will be installed on the CCW system to support flow verifications.
The margin of safety as defined in the basis for any technical specification is not reduced because the CCW system is not required during cold shutdown per Tech-nical Specification 3.4.
SPECIAL TEST T-226 The purpose of this test is to verify that upon loss of instrument air, Contain-ment Isolation Valves CV-0911 and CV-0940 can be closed, and will remain closed for 30 minutes.
SAFETY EVALUATION
SUMMARY
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased by the performance of this test.
The test will be done with the plant in cold shutdown.
Therefore, component cooling water will not be needed for containment equipment.
Air accumulators for CV-0911 and CV-0940 will be tested to see that they provide a back-up to the plant air system.
Upon a loss of air, CV-0911 and CV-0940 fail in the open position permitting cooling flow to CRDM seals, letdown cooling and PCP seal cooling.
In the event of a containment high pressure signal the valves close to provide containment integrity.
A loss of air will allow the control valves to go to the open position.
Therefore accumulators are' needed to close the control valves or to keep the valves in the closed position.
T-226 will test the accumulator function by isolating air to them and.then using stored air to stroke the valves closed.
The accumulator pressure will be taken at the time the control valves are closed and at the end of the test to see if there are any air leaks which would cause the control valves to drift open.
Therefore, successful completion of this test will increase confidence that back-up equipment will perform its.-function i the ~vent of a loss of plant air.
SPECIAL TEST T-234 The purpose of this test is to demonstrate the Reactor Protection System (RPS) response time is more conservative than the response time assumed by the Plant Transient Analysis of the Palisades Reactor for Operations at 2530 MWt.
MI0188-265A-TC07
8 SAFETY EVALUATION
SUMMARY
This test will be performed with the reactor in hot shutdown condition and will consist of inserting two reactor trip signals into the RPS causing a particular logic matrix to deenergize the clutch power supplies.
Each of the six matrix trip ladders will be exercised with 2/4 logic trips from high power, low Steam Generator Number 1 and 2 low level, low Steam Generator 1 and 2 low pressure, pressurizer high pressure and thermal margin/low pressure.
The RPS trip actua-tions will not,differ from that of normal RPS operation except the trips will be manually initiated through the RPS Sensor Instrument Loops.
The RPS will be tripped approximately 150 times to obtain proper response time st~tistical evalu~tion. The RPS rated actuation specification is in excess of 10 trips, therefore the challenge presented to*this system is considered insig-nificant.
Data obtained in this test will be combined with corresponding parameter trans-mitter response time data for determination of overall RPS response time.
The acceptance criteria included in this test will then compare actual RPS time '
response for various trip parameters to those specified in FSAR Section 14.1, and transient analysis XN-NF-77-18.
Therefore, th,e probability of occurrence.
or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased *.
This test will be performed during plant hot shutdown mode to enable the entire test to be performed at one time.
The zero power mode bypass feature of the RPS will riot allow convenient testing at less than hot shutdown condition.
While at hot shutdown,.the plant operating parameters will allow RPS tripping actions and yet require no operatibns attention.
Prerequisites stated in the procedure flag any special safety related or non-safety related actuations or bypasses that are relevant.
Therefore, the possibility of an accident or malfunction of a different type than any previously evaluated in the FSAR is not created.
No technical specification basis are affected through the performance of this test.
The test is being performed to ensure the assumptions made in the transient analysis for 2530 megawatt operation are valid and can support the basis of
. Section 2.3 of the Technical Specifications.
SPECIAL TEST T-256 The purpos~ of this test is to demonstrate the response of Diesel Generator EDC 1-1 to the start of the 'Auxiliary Feedwater Pump P-8A while 'the diesel is supplying the forecast Design Accident ~oad~-
SAFETY EVALUATION
SUMMARY
This test will be conducted on one channel of the plant electrical system in cold shutdown.
The other channel will be unaffected by the test and available in supplying all safety functions including a loss of offsite power.
Existing protective devices will protect equipment from damage due to a sus-tained undervoltage.
MI0188-265A-TC07
9 During certain phases of the test, Switchgear Buses lF and lG will be deenergized.
There will also be outages of Startup Transformers 1-1, 1-2 and 1-3 in prepara-tion for the test and during restoration from the test.
None of these outages will affect the 1-D Bus and its associated safety equipment.
Therefore, the possibility of occurrence or the consequences of*an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The possibility of an accident or malfunction of a different type than any ref er-enced previously. in the FSAR is not created because the worst case failure would result in the loss of one safety related channel of the electrical system which is a scenario that is evaluated in the FSAR.
The margin of safety as defined in the basis for any technical specification is not reduced.
The only limiting condition of operation that will result from this test is one diesel generator being out of service for the duration of the test,
_which will be substantially less than the time allowed by Section 3.7 of the Technical Specifications.
SPECIAL TEST T-258 The purpose of this procedure is to pressurize E-50B with helium (100 psig),
drain E-50B and perform tube leak detection with heliu~ as the testing medium.
Due to the potential consequences for damage to the B Steam Generator, the maintenance of* proper chemical conditions in the secondary side is of extreme
~mportance.
SAFETY EVALUATION
SUMMARY
Helium will be supplied to E-50B via the surface blowdown line and the auxiliary feedwater line.
Pressure will be limited to 100 psi, well below the 350 psi limit established in the technical specifications.
There will be two relief valves in the helium supply set at 150 psi to add overpressure protection.
The steam line hangers have been pinned from the containment penetration to the MSIV CV-0501, plus one hanger downstream of CV-0501.
Hanger pinning has been evaluated and accepted.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The plant is in cold shutdown, no steam is in the system. all components have been evaluated* for acceptability of the cond.itions of this test. Therefore, the possibility of*an accident or malfuuction of a different type than any evaluated previously in the FSAR is not created.
A review of the technical specifications with a detailed review of Section 3.1 as compared to the limits of this test shows that limits will not be exceeded.
A review of the basis also shows this to be true.
Therefore, the margin of safety as defined in the basis for any technical specification is not reduced.
MI0188-265A-TC07
- e.
10 SPECIAL TEST T-259 The purpose of this test is to establish a loop seal on the inlet to CV-0501 to ensure it can hold a seal for subsequeni E-50B nitrogen or helium pressurization.
The seal will be established and then held for the subsequent tests.
SAFETY EVALUATION
SUMMARY
The steam line hangers have been pinned from containment penetration to one hanger past CV-0501.
The plant is in cold shutdown.
There is level indication provided and a procedural limit to maintain water level below that analyzed in EA-SG-TMP-1, that level being 5 feet vertical from the penetrati'on to containment.
Hanger pinning has been evaluated and deemed acceptable per EA-SG-TMP-1.
There-fore, the probability of occurrence or the consequences of an accident or mal-function of equipment important to safety previously evaluated in the FSAR is not increased.
Because the plant is in cold shutdown, no steam is in the system, the components have been evaluated for acceptability of the conditions of this test. Therefore,,
the possibility of an accident or malfunction of a different type than any eval,-
uated previously in the FSAR is not c_reated.
A _review of the technical specifications with a detailed review of Section 3.1 shows that the margin of safety as defined in the basis for any technical speci-fication is not reduced. *
- SPECIAL TEST T-249 The purpose of this test is to test the operability of HPSI Redundant High Pressure Safety Injection and Hot Leg Injection motor operated valves.
The testing is performed in response to IE Bulletin 85-03.
NOTE: One train of the HPSI _system was tested in 1987.
SAFETY EVALUATION
SUMMARY
This procedure is to be performed as part of our response to IE Bulletin 85-03, Motor-Operated Valve common mode failures during plant transients due to improper switch settings.
During this test the stroke time operability of M0-3007, 3009, 3011, 30-13, 3062, 3064, 3066, 3068,* 3080, 3081, 3082 and_ 3083 will be verified during the maximum expected differentiai pr~ssure that the valves would be *sus-ceptible to during the designed acciden,t conditions _listed in FSAR Section 6.1.
During the performance of this test the operating limits of the system will be within the bounds of the accident conditions in FSAR 6.1.
Therefore the proba-bility of occurrence of an accident or malfunction is not increased.
This procedure is used to verify that specific High Pressure Safety Injection, Redundant High Pressure Safety Injection, and Hot Leg Injection limitorque operators will actuate in the required time under the condition of maximum differential pressure across the valve disc during design accident conditions.
Also, since the maximum differential pressure is within the operator's boundary MI0188-265A-TC07
ee specified in FSAR Section 6.1, the possibility of an accident or malfunction not previously evaluated in the FSAR is not increased.
This test will be performed during cold shutdown conditions and therefore technical specifications are not affected.
SPECIAL TEST T-250 11 The purpose of this test is to test the operability of the Auxiliary Feedwater Motor Operated Valves in a post-accident situation.
This testing is in response to IE Bulletin 85-03.
SAFETY EVALUATION
SUMMARY
- This procedure is to be performed as part of a response to IE Bulletin 85-03, motor operated valve common mode failure during plant transients due to improper switch settings.
During this test the stroke time operability of M0-0743, 0749, 0753, 0754, 0755, 0759, 0760 and 0798 will be verified during the maximum expect-ed differential pressure that the valves would be susceptible to during the designated accident conditions listed in the FSAR Section.9.7.
During perform-ance. of this test, the operating limits of the system will be within the bounds of the accident analysis listed in FSAR Section 9.7. Therefore, the probability of occurrence or the consequences of an accident or malfunction of equip~ent important to safety previously evaluated in the FSAR is not increased.
This procedure is used to verify that specific auxiliary f eedwater limitorque operators will actuate in the required time under the conditions of maximum differential pressure across the valve disc during design accident conditions.
Since the maximum differential pressure is within the operating boundary speci-fied in FSAR Section 9.7, the possibility of any accident or malfunction of a different type than any evaluated previously in the FSAR.is not increased.
This test,will be performed during cold shutdown conditions and therefore tech-nical specifications 3.5.l and 3.5.2 are not affected, and the margin of safety as defined in the basis for any technical specification is not reduced.
MI0188-265A-TC07
ee MI0188-26SA-TC07 OPEN FACILITY CHANGES AND SPECIFICATION CHANGES INITIATED IN 1987 REPORTABLE ON THE ANNUAL REPORT NOT DECLARED OPERABLE OR CLOSED OUT 12
13 SC-87-204 This specification change involves the elimination of 15 hydraulic and three mechanical snubbers.
Two of the mechanical snubbers are located on the Feed-water System and are not safety ~elated.
SAFETY EVALUATION
SUMMARY
Stress packages containin~ the snubbers were modeled in ADLPIPE and were bench-marked against the original Bechtel results. Palisades Plant Technical Sp.ecifi-cation M-195 was used for design requirements and analysis criteria.
This specification defines the design and analysis criteria of the updated FSAR.
By dynamic analysis of the modeled stress package, it was shown that five snubbers and their support locations could be totally eliminated.
By static thermal analysis, it was shown that the remaining 12 snubbers covered by the specification change could be replaced by equivalent capacity struts.
The thermal analysis with a strut in place of the snubber.demonstrated that the resulting loads and stresses are acceptable.
Amendment 107 to the Palisades Operating License deleted Tables 3.20.1 and 3.20.0 containing the list of snubbers at Palisades from the technical.speci-fications.
As stated in the Safety Evaluation related to Amendment 107, Consumers Power Company agreed that.the identification of the applicable snubbers will be added to the updated FSAR.
FSAR*Change Request 5-13-R4-159 was completed November 16, l987 which added Tables 5.7-10 and 5.7-11 containing the list of snubbers to the FSAR.
These tables will be revised to reflect the snubbers removed as a result of this specification change.
Snubbers listed in the FSAR will be removed~ but these modifications are being performed using design and analysis criteria given in the FSAR.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety has not been increased.
The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created because the design and analysis criteria used for this specification change is that given in the FSAR.
The technical specification and testing requirements for snubbers are based on a percentage of the number of safety related snubbers installed in the plant.
Removal of snubbers does not reduce the margin of safety.as defined in the technical specification because the same percentage of safety related snubbers will be inspected and tested.
SC-87-332 This specification change covers changing the LTOP alarm from 325 psia to a multiple LTOP alarm and actuation setpoint, to be calibrated as PCS temperature changes during heatup and cooldwon, utilizing two setpoints during heatup, and two setpoints for cooldown.
MI0188-265A-TC07
14 The previous single LTOP actuation setpoint at 375 psia was determined in 1977.
Due to recent revisions to Appendix G heatup/cooldown limit curves, the 375 psia setpoint no longer adequately protects the reactor vessel for all "Anticipated Operational Occur.rences" postulated in 1981 in Branch Technical Position RSB-5-2 Revision 0, dated 7/81.
SAFETY EVALUArION
SUMMARY
The accident considered is that of exce*eding the allowable pressure versus temperature stress curves for the reactor vessel per Branch Technical Position, RSB 5-2, specific overpressure initiating events, or "Anticipated Operational.
Occurrences" to be considered for LTOP analysis are:
- 1.
A charging/letdown system imbalance
- 2.
Inadvertent start of HPSI pump
- 3.
Starting of primary coolant pump with the secondary side hotter than the primary side.
Item 1 has been analyzed *by the Accident and Transient Analysis Group for the extreme case of three charging pumps in service, heatup rate of 90°F/hour, and complete letdown failure. It is demonstrated in EA-E-PAL-85-101 that the pro~
posed LTOP multiple setpoints will protect from exceeding Appendix G allowable curves*in this "AOO" for both "PCS Solid" and "Pressurizer Bubble" conditions.
Item 2 is precluded from occurring by administratively removing HPSI pump control fuses below 430°F.
At 430°F, the vessel is not changed by the full shutoff head of HPSI pumps.
Item 3 is precluded from occurring by administrative controls.
A primary coolant pump shall not.be started with the secondary hotter than the primary side per SOP-1.
Thus, the probability of occurrence or the consequences of an accident* or mal-function of equipment important to safety prev.iously evaluated in the FSAR is not increased.
The proposed setpoint changes for LTOP are acceptable, not exceeding reactor vessel pressure versus temperature limits.
The possibility of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created.
The margin of safety for reactor vessel ver~µs pressure limits are inherent in the Appendix G curves.
The proposed seipoint changes have been shown to protect from exceeding these curves.
Therefore, the margin of safety as defined in the basis for any cechriical specification is not reduced.
SC-87-350 & SC-87-351 These specification changes cover the tube plugging in A and B Steam Generators to remove tubes from service which exceed the Palisades plugging criteria as specified in the 1987 Steam Generator Eddy Current Test Outage Plan, and return the steam generators to an operable condition.
MI0188-265A-TC07
15 SAFETY EVALUATION
SUMMARY
The tubes to be removed from service are those which have exceeded Palisades plugging criteria as specified in the 1987 Steam Generator Eddy Current Test Outag~ Plan.
By removing the d~fective degraded tubes from service, the possi-bility of a primary to secondary leak will greatly decrease.
The total number of tubes plugged does not exceed the amount which would require plant derate to maintain the integrity of the fuel.
Qualified plugs and qualified mainte-nance and welding procedures are to be used ensuring that correct practices and materials will be utilized, reducing the potential for a tube plug weld failure.
Tests will be performed as required by Procedure PCS-M-42/43, and Technical Specifications 4.4 and 4.15 to ensure plug and system integrity as well as PCS flow.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
A detailed review of FSAR Sections l.*l, 1.2.4, 1.5, 3.0, 4.0, 6.9, 14.15, 14.22 and Tables 1-2 and 4-4 indicates that the probability of an accident or malfunction of a different type than any evaluated previously in the FSAR is not created. '
A detailed ieview of technical specifications was conducted.
Specification 4.15 requires a PCS flow measurement after 10 or more tubes are plugged to verify flow against the basis for Specification 3.1.l(c) to ensure the tube plugging will not affect fuel integrity.
Specification 4.14 delineates the requ~rements of the Augmented Inservice Inspection Program for Steam Generator.
The basis statement identifies that Palisades has not observed any tube leakage since June 1974.
Since this is no longer true, the technical specifications will be changed.
However, this statement has no effect on the Engineering Basis, Calcu-lations, or Margin(s) of safety as identified in Specification 4.14.
Specification 4.4 requires a surface examination as well as a system leak test after this maintenance to ensure system integrity, and leak tightness during normal operation.
Based on the above, it can be concluded that after the ident-ified testing is complete, the margin of safety as defined in the basis for any technical specifications is not reduced.
FC-731 This facility change covers the transmitter-replacement per Regulatory Guide 1.97 as follows:
- 1.
Upgrade the HPSI and LPSI flow indication instrumentation loops to meet the Category 2 requirements of Regulatory Guide 1.97.
This wiil be done by in-stalling IEEE Class lE, environmentally qualified, safety-related transmitters, cables and containment penetration connectors; five valve manifolds, square root extractors and linear scale indicators for the following loops:
HPSI - FT-0308, FT-0310, FT-0312, FT-0313 LPSI - FT-0307, FT-0309, FT-0311, FT-0314 MI0188-265A-TC07
16
- 2.
Upgrade the pressurizer wide range level indication loop (LT-0102A and LI-0102A) to meet Category 1 requirements of Regulatory Guide 1.97.
This will be done by installing IEEE Class lE, environmentally qualified, safety-related environmentally qualified, safety-related transmitter, cable and
- containment penetration connector; five valve manifold, and linear scale indicator with a high level alarm setpoint.
- 3.
Removal of the Pressurizer Level Indication Loops LT-0102B, LT-0102C and LT-0102D along with their associated indicators.
This includes removal of cable, conduit, instrument tubing and installation of blanks in the panels where the indicators were.
Note: LI-010.2E indicator is replaced under SC-87-032.
Both the flow transmitters.and the level transmitters will be fitted with environmentally qualified seals.
SAFETY EVALUATION
SUMMARY
The operation of the nine instrument loops will be unchanged.
The loops are beinl!1a upgraded via replacement of the in-containment loop components (trarismitters, cables, penetration connectors, splices and conduit seals) with functionally equivalent ones that are IEEE Class 1E and environmentally qualified.
Out-of-containment components are being replaced with those that are lE.
Once.upgraded, these loops will be more reliable.
Indicators will fail low to prevent reading
,of a failed indicator as a valid flow or level.
Use of square root extractors in the HPSI and L_PSI flow loops, along with linear scale indicators will provide better resolution at low flow rates.
This facility change also removes LT/LI-0102B, LT/LI-0102C and LT/LI-0102D.
These are redundant LT/LI's to LT/LI-0102A.
Suf-ficient redundan~ level instrumentation is provided by LT-0101A, LT-0101B, LT-0101C and LT-0101D v*ia LIC-0101A and LIC-0101B and LT-0103 so that the prob-ability or conseque~ces of an accident or malfunc~ion of equipment important to safety previously evaluated in the FSAR are not increased.
The level transmitter in the pressurizer level loop LT/LI-0103 is a fully qualified device.
Right and left channel level indication will be provided by Loops LT/LIA-0102 and LT/LI-0103.
The possibility of an.accident or malfunction of a different type than any eval-uated previously in the FSAR is not created because the new loops serve the same function as those they are replacing.
The marg-in of safety as defined in the basis for any technical specification is increased by the ins.tallation of these qualified components.
A revision to.the technical specifications is necessari; Section 3.17,* 4-8 to reflect the removal of the wide range.
Pressurizer level indication loops LT/LI-0102B, LT/LI-0102C and LT/LI-0102D, and the incorporation of loop LT/LI-0103 as redundant indication.
The LT-0102B, LT-0102C and LT-Q.102D transmitters are not included in the Basis Statement 3.17.2 of the technical specification.
Removal of these loops will reduce the likelihood of conflicting indications.
MI0188-265A-TC07
17 FC-735 Although this modification does not meet the criteria of being included into the Annual Report, it is being included because it is an NRC commitment.
This modification covers the permanent installation of P-66A and P-66B HPSI Pump high accuracy discharge pressure test gauges to permit more accurate determina-tion of HPSI pump discharge pressures.
SAFETY EVALUATION
SUMMARY
The pressure gauges will be connected to the HPSI Pump discharge root isolation valves whicp are normally closed via tubing.
Per Regulatory Guide 1.26, all piping and associated equipment downstream of the root valve is non-class.
How-ever, since the root valve is Class II", per Regulatory Guide 1. 26, material com-patability is a concern.
Therefore, materials will be procured in accordance with existing specifications to ensure the integrity of the root valve/pump dis-charge is not.compromised.
The installation is judged not to present a seismic concern:as the gauge will be affixed to the building via a stanchion.
Should this mounting fail allowing the equipment to fall, there is no equipment important, to safety immediately below that could be damaged.
The tubing weight added to the piping system is not significant to degrade the integrity of the piping during a seismic event.
Since the materials are consistent ~ith existing specifications, the installation will be performed in accordance with applicable codes, and the modification will not degrade the system.
The probability or consequences of an accident will not be increased.
The modification will not change the system function or operation.
tion will be in accordance with existing codes and specifications.
possibility of an accident or malfunction of a different type than previously in the FSAR is not created.
The installa-Therefore, the any evaluated*
Since the system function is not changed, the margin of safety as defined in the technical. specification is not reduced.
FC-737 This facility change has been initiated for the replacement of DBA/NSD Sequencers.
The existing sequencers will be replaced with a programmable controller (PC).
The PC will perform the exact function that the ~xisting DBA/NSD Sequencers perform.
They will be seismically tested, and mo~nted in the same place that the existing sequencers are (in Panel C-13).
Each channel (right and left) of the safety injection and sequence loading circuit (S41 and S42) now contains two DBA sequencers and one NSD sequencer.
They are Eagle Signal HO series sequencers and have approximately 30 loads divided among the three.
MI0188-265A-TC07
- ,.I 18 This modification will replace the three sequencers in each channel with one PC.
The exact same loads from the three sequencers will now be sequenced by the PC.
The timing of this sequence will be changed to take advantage of the better timing accuracies of *the PC.
Terminal boards will be added (space permitting) to allow for easier removal from service of the PC.
The sequencer circuitry wil~ be changed to allow for the PC to perform the resetting function.
SAFETY EVALUATION
SUMMARY
This modification will replace the existing NSD and DBA Sequencers with a pro-granunable contr9ller (PC).
This replacement is being made to preclude the mechanical failures which have been experienced.with the existing sequencers.
This evaluation will discuss the three changes to be made in this modification.
They are: 1) Replace the existing three sequencers in each channel with one PC.
- 2) Change the start times for some of the loads.
- 3) Change the control circuitry*
for sequencer resetting.
The sequencers are required to operate in order.to sequence loads onto the diesel g~nerators.
-T~is sequencing is required to contend With various postulated accident scenarios.
The worst case sequence is s~own in the FSAR Tables 8.6 and 8.7.
There are two redundant (left and right) channels of load sequencing and ~ach channel presently contains three sequencers.
Two of the three sequencers are required for the design basis accident (DBA) due to the large number of loads being sequenced in this scenario, and one sequencer is dedicated to the normal shutdown (NSD) sequence.
The DBA.sequencers are activated on a safety injection signal (SIS) coincident with a loss of standby power.
The NSD sequencer is
.activated upon loss of standby power alone.
Mechanically, each sequencer currently consists of a motor which drives a long cylindrical shaft.
Each shaft is approximately three feet long in the case of a DBA Sequencer and one foot in the case of the NSD Sequencer.
In each case, the shaft contains cams aligned on the shaft in various positions to create different starting times for each sequenced emergency load.
Item 1.
In* total, two PC's are to be installed. *Each PC will perform both DBA and NSD *sequencing operation for its designed channel.
These two PC's will be installed in the control cabinet slots vacated by their predecessors~ The two PC's will be physically separated and electrically independent, thereby meeting the requiremenj;s_ of IEEE 279-1979 and the intent of IEEE 384-1977.
The replacement of the existing sequencers with a PC in each channel will provide more reliable and accurate hardware to perform the sequencing function.
Industry data indicates that the PC is highly reliable.
The PC will be seismically anchored and will be tested.by CPCo to b.e seismically functional per IEEE 344-1975.
The combination of both the NSD and DBA sequencers into on PC hardware package for each sequencing channel will have no effect on the probability of occurrence MI0188-265A-TC07
19 or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR.
Each PC hardware package will utilize one input to initiate the DBA sequencing outputs and another input to initiate the NSD sequencing outputs.
This is the same way that the present circuit is set up using three pieces of hardware.
A total failure of a programmable controller is not likely.
The PC Processor, power supply, and input signal will all be energized from a preferred AC power panel as the existing sequencer.
The output signals will be wired ~n the same manner as currently exists.
This is with the exception of the loads on the NSD sequencers.
Currently, these loads are activated by either a DBA or NSD sequencer contact (wired in parallel) depending on the initiating scenario; each contact set to close at its own prescribed time interval.
In the new sequencer, only one output per load will be used.
This same output (contact closure) will be utilized for either initiating condition; ie, DBA or NSD.
The correct input representing the initiating condition (DBA or NSD) will determine which sequence will be used for the load and will close the load-initiating contact at the prescribed time for that condition.
Another exception is the elimination of the redundant sequencer output contact which was installed in the spring of 1987 to increase the reliability that a single contact failure on the sequencer would not adversely affect the sequence.
This contact failure and most all past sequencer failures have been mechanical in nature; that is, a failure in the cams, shafts, or gears, which all must physically move in harmony to start loads in the prescribed sequence.
The.PC is an electronic component and will not be susceptible to these mechanical fail-ures.
Therefore, the double contact on a load set-up is no longer necessary.
Item 2.
This item will further eliminate the potential for overlap in the start-ing of adjacent sequential loads.
Any new load start times will be bounded by the values given in FSAR Tables 8.6 and 8.7 or the values used for the accident analysis as shown in Chapter 14 of the FSAR.
Item 3.
The control circuit changes will be made to allow sequence resetting to be performed by the PC.
Presently, the resetting is performed by operators.
No change in function of any engineered safeguards component will result from this
- circuit change, as all actuated loads will have sequenced on and resetting the sequencer will not trip any load breakers.
Additionally, the following backup systems will be available on the PC w~ich are not provid~~ on the present sequencer.
The PC has an internal battery which would supply power to maintain the processors memory in the event of a loss of power to the PC.
The status of the battery wilL-be administratively controlled so that it does not becom!. ~ischarged.
The PC also has an attachment called an EE Prom.
This is a memory chip which will load the sequence memory into the PC in the event that normal power is lost to the PC and the battery is dead.
The EE Prom needs no electricity to operate.
The PC will constantly be monitoring itself for internal faults.
If it detects a fault it will activate an annunciator in C-13 Control Panel.
MI0188-265A-TC07
20 As shown above, the PC is more reliable, is not prone to mechanical failures and has backup systems to connect with a loss of. both power sources (normal and bat-tery).
Failure of the PC to correctly perform the sequencing function is highiy unlikely. If an internal fault is present in the PC, it will likely be identified by the PC and annunciated to the control operators. '
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
The possibility of an accident or malfunction of a different type than any eval-
.. uated previously in the FSAR is not created because.an internal fault on the PC
- will be detected and annunciated to the control operators.
Action will then be taken to repair the PC.
It is not considered credible that both sequencers would fail at the same time, as the design meets the single failure criteria. If failure of one PC were to occur, the operators would have to manually start the loads.
Failure of the PC will not block the control circuit such that the operators could not start the loads manually from the control room.
The PC will be seismically mounted and tested.by--CPCo for ability to operate during and after a seismic event.
The margin of safety as defined in the basis for technical specifications has not been reduced because the PC is more* reliable than the existing sequencer.
It _also has ba.ckup features not available on the present sequencer.
FC-755 This modification covers the replacement of the C-47 Fire Protect.ion Deluge Panel.
The existing C-47 panel has 24 alarm windows for the deluge fire system.
However,.
the master alarm from this panel to the control room panel C-13 does not have a reflash feature.
The new C-47 panel will have 36 windows and it will have reflash alarm capability.
The new C-4 7 panel will be located in the s.ame control room position as the old panel.
The new master file alarm from Ultra Violet Detector. Panel C-132 in the intake structure will be wired to C-47 via J470.
The existing fire alarm from the support
- building will be added to C-4 7.
The existing EER Room water flow switch fire alarm which is presently wired in parallel with the lD Switchgear Room fi'I'e alarm will be.._wired as separate alarms to ia new junct!o~ box J470, and then to C-47 Panel.
The new panel and all alarms will 'be tested prior to return to service.
This modif fcation will resolve the problem of reflash capabilities described in Event Report E-PAL-82-143.
MI0188-265A-TC07
SAFETY EVALUATION
SUMMARY
This project has been evaluated "Q" due to Technical Specification Limiting Conditions of Operation Section 3.22.1, and seismic ~ounting of Panel C-47, new Junction Box J47D, and related conduits.
The new (E)C-47 panel will be mounted in the control room in the same place as the existing panel and the new junction box wil.l be mounted on a wall in the cable spreading room, near
.the battery room.
21 Construction will use both approved anchor bolts and the proper method of in-stallation to prevent degrading the seismic,criteria of the Auxiliary Building walls.
By following the requirements listed above, we have not increased the probabil-ity of occurrence or the consequences of an accident or malfunction of equipment important to* safety previously evaluated in the FSAR.
The possibility of an accident or malfunction of a different type than any eval-uated in the FSAR is not created since the.new C-47 Panel performs a similar function to the old panel.
The margin of safety as defined in the basis for any technical specifications is not reduced because some existing fire alarms that are now in parallel with other alarms will be separate alarms which should provide operators with more information in the control room.
This replacement panel does not violate any of the technical specification require-ments, except that fire watches wil1 be required during construction.
. This facility change covered the computer modifications to support rod drop timing.
By letter to the NRC dated May 5, 1986, Palisades committed to providing the capability of recording control rod positions for use during post-trip evalua-tions.
This change modifies the Primary Information Processor (PIP) and Secondary Position Indication (SP!) computers to allow them to communicate information on control rod.. position to the Critical Function Monitor System (CFMS) computer.
The CFMS*ii capable of recording rod positions with sufficient resolution to determine if rods dropped into the core, or.were driven into the core, following a reactor trip.
This change thus satisfies the commitm~nt to the NRC on rod drop timing.
This change consists of the following items:
- 1.
Modify the PIP and SPI computers to increase memory size and provide an interface for communications to the CFMS computer.
- 2.
Provide cabling betwe*en PIP and. CFMS and between SPI and CFMS.
MI0188-265A-TC07
e*
- 3.
Modify PIP and SPI software to allow sending data to the CFMS.
- 4.
Modify CFMS software to accept data from SP! and PIP and to display data on appropriate display pages.
SAFETY EVALUATION
SUMMARY
22 The function of the PIP, the SP! and the CFMS are described in FSAR Sections 7.6.1.3, 7.6.1.6, 7.6.2.3 and 7.6.2.6.
These computer systems function to monitor display and alarm plant data.
In addition, the PIP provides interlocks for control of the control. rods.
Failure modes of these computer systems are such that any malfunction will not result in the occurrence of an accident or prevent the operation of required safety systems..The systems are classified as non-Class lE and no reliance on the operation of these systems is taken credit for in the safety analysis.
This facility change adds new solid state memory to the PIP and SP! computer to replace obsolete core memqry and interfaces the PIP and SP! computer to the CFMS computer to allow the CFMS computer to monitor control rod positions.
No existing functions presently performed by any of the computer systems are deleted by this modification.
The modification does not result in any new failure modes of the computer systems which could effect the probability or consequences of an accident.
The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is npt increased.
The modification will not result in any changes in the effect on the plant due to failure of any of the effected computer systems, thus the possibility of an accident or malfunction of a different type ~han previously evaluated in the FSAR is not created.
The PIP and SP! systems are utilized as monitoring devices to assure conformance to technical specification requirements on control rod positions and power dis-tr~bution. Modifications to these computers are made by this facility change do not' affect the capability to monitor the parameters required by these technical specifications.
Thus, the margin of safety as defined in the basis of the tech-nical specifications is not reduced.
FC-774 This modification covered the removal of Radiation Monitors in the Feedwater Purity Building, including 7 area monit-0rs and one proc~ss monitor - exempting them from any !e~hnical specification and/or administrative requirements.
This modification also removes the dirty waste sump pumps P-940A and P-940B pump start interlock on RIA-8265 radiation alert signal.
SAFETY EVALUATION
SUMMARY
This modification does not affect nuclear safety in any way.
Area and process monitors. in the Feedwater Water Purity Building serve no useful purpose, since MI0188-265A-TC07
---*~
23 the condensate demineralizer system is no longer used.
Radiation levels in the building should not increase enough to warrant radiation monitoring instrument-ation.
Therefore, the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
Condensate flow is bypassed around the Feedwater Purity Demineralizer System.
Therefore, the possibility of contaminated condensate circulating around the Feedwater Purity Building is removed.
Contaminated condensate would be de~
tected by* main steam line monitors and off-gas monitors.
Thus the possibility of an accident or malfunction of a different type than any evaluated in the FSAR is not created.
The margin of safety as defined in the basis for any technical specification is not reduced because the technical specifications do not include a basis or LCO involving plant alarm monitors.
MI0188-265A-TC07
consumers Power POWEIUNli lllllCHlliAN'S PIUlliRESS General Offices: 1946 West Parnell Road, Jackson, Ml 49201 * (517) 788-0550 March 31, 1988 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-255 - LI_CENSE DPR PALISADES P~T -
1987 ANNUAL REPORT OF FACILITY CHANGES, TESTS AND EXPERIMENTS Attached is Consumers Power Company's Annual Report describing the Facility Changes (FC), Specification Changes (SC), tests and experiments initiated at the Palisades Plant during 1987 whether_ or not they were declared operable.
Exceptions to this statement are as follows:
SC-83-138, SC-86-323, FC-571 and FC-601.were initiated prior to 1987 and were not previously reported on the annual. rep()rt.
They were declared operable and closed out in 1987.
FC-683 was closed. out in 1986 and not reported on any previous annual. report.
Although FC-683 is not reportable under 10CFRS0.59, it is being included in the 1988 Annual Report because of its involvement in an FSAR change that was documented in Consumers Power Company letter dated October 30, 1987 regarding revisions to the FSAR.'
This report is submitted in accordance with the provisions of 10CFR50.59(b).
Brian D Johnson Staff LicensJ~g Engineer CC Administ~a~or, Region III, NRC NRC Resident Inspector - Palisades Attachment OC0388-0078-NL02
'