ML20042F761
| ML20042F761 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 10/31/1989 |
| From: | Tucker H DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9005090360 | |
| Download: ML20042F761 (250) | |
Text
--
- l
- Duke 1%uer Company lia R k ur f
PO Bar33196 Charlotte, N C 28242 l' ice President Nuclear Pmductwn Y'
(701)373-4531 DUKE POWER ,
h May 1, 1990 ,
L L
U. S. Nuclear Regulatory Commission A'lTN - Do'cument' Control Desk Washington, D. C '20555 Subjoet Catawba Nuclear Station.-Units 1 and 2 ,
Docket Nos. 50-413 and 50-414 1989 10 CFR 50.59' Report Gent.lemen: .
Pursuant to 10 CFR 50.59, please find attached a summary of Nuclear Station .
Modifications, Exempt Change Variation Not. ices, procedure chanRes, tests.
-and experiments which were completed under t.he provisions of 10 CFR 50.59 from January 1, 1989 to October 31, 1989. j a
Very truly yours.
ara k /
Ital B.' Tucker Milli-35/lcs ' -
i xc: Mr. S. D. Ebneter Regional Administrator, Region II.
U. S. Nuclear Regulatory Commission l' 101 Marietta St., NW., Suite 2900 Atlanta, Georgla .30323 Mr. W T. Orders NRC Resid(nt Inspector ,
Catawba Nuclear Station ,
T 1
9005090360 891031 s PDR ADOCK0500g3
.a Ie 47 ;
- i. 1 Catawba Nuclear Station Summary of Nuclear Station Modifications Completed From 1/1/89 - 10/31/89 CN-00006
Description:
This modification provides larger, more convenient areas for the storage of tools in contaminated areas. This l f- modification will eliminate the need to go up a flight of steps-to I access-the hot tool crib. The new hot tool crib will have the following features: A second floor designed to 300 psf, stairs to access second level, a 1/2 ton hoist'for handling materials, l storage areas, issue counter, telephone, adequate ventilation for j a contaminated area, floor drains connected to the radwaste i system. This modification affects FSAR Figures 9.2.3-2, 9.3.1-7, p
9.5.1-6, 11.2.2-15.
Safety Evaluation: No safety system will be degraded, and no functional change will be made to any system as a result of this modification. The enlarged hot tool crib will reduce the need to L lift heavy tools up a flight of stairs and will be convenient to more areas of the station than the existing tool crib. No unreviewed safety question is Juaged to be created by this modification.
CN-10097
Description:
This modification installs an oil drain line and '
valve for each of the upper and lower lubricatiots oil access ,
, ' ions of the four RN pump motors. This change affects FSAR
- "s 9.2.1-1 and 9,2.1-3.
Evaluation: A sufficient method was not provided for
.. m ining the lubrit.ating oil of the four Nuclear Service Water (RN) pump motors. This modification will install tubing on the 4 RN pump motors at the upper and lower lubricating oil access locations. The tubing will provide a method to maintain the lubricating oil. This modification will not degrade the RN System. It will make the RN pumps more reliable by making it possible to replace the oil for the pump motor.
This modification will not increase the probability or conse-quences of any accidents previously evaluated in the FSAR and no
- new accidents are introduced. Applicable design practices are being -followed for the piping, therefore; the probability or consequence of any equipment malfunction important to safety will not be increased. No new failure modes have been determined which will increase the likelihood of radioactive releases outside of containment. The margin of safety as defined in the bases of the Technical- Specifications has not been reduced by these NSM(s). l There are no unreviewed safety questions associated with these NSM(s).
CN-10144
Description:
This modification provides a reliable back-up power L source for radiation monitors 1 EMF 26, 1 EMF 27, 1 EMF 28, and 1 EMF 29.
l l
2 Safety Evaluation: There is presently no backup power for the steam line area radiation monitors. Operability of these monitors is required by regulatory commitment in the Catawba FSAR,. Table 12.3.4-1 Offsite Dose Assessment, Modification will enhance the survivability of this system and will not adversely impact any safety or non-safety system. Appendix R fire protection criteria have been considered.
CN-10412
Description:
This modsfication will alter the way that IEMF-31 gets flushed. Presently 1 EMF-31 is flushed in the same direction that its process fluid flows and the water for this flushing is provided by conventional low pressure service water (RL). Upon completion of this modification IEMF-31 will be back flushed with water from the demineralized water (YM) System.
Safety Evaluation: The functional capabilities of IEMF-31 which are to provide radiation information about the Turbine Building.
Sump and based upon that information control the discharge path of the sump will be the same following completion of the modifi-cation. 1 EMF-31 is a non QA condition component and does not perform a role in mitigating any design basis accidents; No plant parameter will be affected by this modification. The YM system will be isolated by a check valve from 1 EMF-31 to prevent contamination of the YM system. Since this modification has no interaction with equipment, structures or components important to ,
safety and does not create a new release path, there is no increase in the probability of accidents of malfunctions of equipment important to safety previously evaluated in the FSAR.
Since all components modified are not degraded, there is no increase in the consequence of accidents or malfunction of equipment important to safety already evaluated in the FSAR, For both of the above reasons, no possibility is created of accidents
.or malfunctions of equipment important to safety different from any already evaluated in the FSAR. No plant parameter or setpoints are altered by this NSM so no margin of safety as ,
defined in the basis to any Tech. Spec, is reduced.
CN-10608 .
Description:
This modification revises control circuits for valves 1N19A, IN1108, 1KC56A, 1KC81B, INF228A, INF233B, INF234A, 1NM3A, INM7B, INM22A, INM26B, INW8A, and 1NW61B to block these valves from repositioning when an S.I. signal is received with control at the Auxiliary Shutdown Panel. This change affects FSAR Tables 7.4.7-1 and 7.4.7-2, Safety Evaluation: This modification will provide Auxiliary Shutdown Panel control of valves that are currently subject to erroneous positioning on receipt of a spurious safety injection signal. Pumps discharge to the NC System.
The blocktrg of S.I. signals to the following valves while at the Auxiliary Shutdown Panel will prevent the described adverse affects:
L J
T1 g
3 (1) NI-9A NV Pumps Discharge Isolation valves to NC cold legs NI-108 (open on SI), Result is rapid filling of PZR by providing an alternate low resistance flowpath from NV Pumps Discharge to the NC System. The operator has no indication or control of the valves from the ASP's and no indication of flow to the NC loops through that flowpath.
-(2) KC-56A A and B train KC inlet isolation valves to the ND L KC-81B Hx's (open on SI). Result is runout flow on inservice KC Pumps since idle KC Pumps do not auto start on SI. Operations does not feel the operator has sufficient time to recognize the excess flow
.and start additional KC Pumps to prevent pump damage.
(3) NF-228A Ice Condenser Glycol Supply and Return Containment NF-233B ~Isolations (close on SI). Result is glycol spill NF-234A inside containment from overflow of glycol expansion tank.
(4) NM-3A NC Loop A sample and PZR liquid sample containment NM-78 isolations (close on SI). Result is loss of NM-22A ability to sample for boron concentration during NM-26B plant cooldown. Boron concentration is used to determine shutdown margin, verify borativn results and to verify PZR and NC System boron concentra-tions are approximately equal.
(5) NW System Containment Penetration Valve Injection Water System (actuates on SI). Result is injection into possibly open valves (since many valves are blocked from SI at the ASP's). Continued injection could result in RN auto makeup to the NW surge tank with subsequent injection of'RN water-into open valves.
This modification has no adverse impact on any safety or non-safety system. No unreviewed safety question-is created by
.this modification.
CN-10733
Description:
Fill and vent valve assemblies will be installed in the RVLIS piping to the Reactor Head to facilitate "Backf1111ng" the upper head sensor piping as required when returning the system to normal line up following reconfiguration for mode 6 or draining and refilling'the Reactor Coolant System.
' Safety Evaluation:
CN-10849
Description:
This modification will install a radiation monitor (EMF-60) for the purpose of detecting the presence of noble gasses in the Annulus. The monitor is to be located in the Auxiliary Building (Aux Bldg) and connected to a sample line that draws annulus air samples from four (4) points within the Annulus. No
- connections to the Annulus Ventilation (VE) system are required.
4 This sample line will have a manual valve and solenoid valve at each sample point. Additionally, three solenoid valves are-located in the Aux. Bldg., one each for inlet and outlet isolation and one for Aux. Bldg. air purge of the sample line. The sample return line is routed back to the Annulus. The radiation monitor 1 air supply line will be heat traced to prevent moisture condensation.
Safety-Evaluation: EMF-60 and the associated piping are non safety related. The components of this sampling system are not required to be seismically qualified but have been evaluated for potential seismic interaction. The piping associated with this NSM will only penetrate into the Annulus and not into Containment and therefore containment isolation will not be required. All radiation released from Annulus will still be filtered by the VE system before release to the environment.
EMF-60's drawing out of air samples from the Annulus and returning the sample back to the Annulus will have no impact on the ability of the VE system to maintain a negative pressure in the Annulus ;
following an accident. Also, the pipe break failure mode of this -
addition will not prevent the VE system from creating a negative pressure. The reason for the lack of impact is due to the vast j differences in the air handling capabilities (VE system 9000 scfm, ;
EMF-60 2-5 scfm) (Ref 1&3). j Since the sample line has the possibility of carrying highly j radioactive air from the Annulus to the EMF which is in the Aux, l Bldg., an automatic isolation / purge feature (activated on sensing i nigh radiation in the sampling air) will isolate the inlet to the i EMF and then purge the sample line with Aux Bldg. air. The :
radioactive air will be purged into the Annulus (Ref-4). This i feature of the sampling system will avoid concern of high j radiation exposure to equipment in the area of the EMF and the :
Aux. Bldg. portion of the process line. Health Physics and procedures will control the use of the sampling system in a bypass mode which permits use of the EMF even though a high radiation signal is actuated. These controls are to avoid equipment radiation exposure concerns. Since no system or piece of equipment is adversely affected by the installation of EMF-60, the probability or consequence of an accident previously evaluated in :
the FSAR is not increased nor is the probability or consequence of j a malfunction of equipment important to safety previously evaluated in the FSAR increased, t
The nature of the EMF and the sampling lines are such that it precludes its potential role as an event initiator. Therefore, the possibility of an unanalyzed accident is not created by the installation of the EMF. l The installation of this NSM will have no adverse impact on any piece of equipment; therefore, it does not create the possibility of an unanalyzed malfunction of equipment important to safety.
l
o si 2
Since this NSM involves only the addition of a monitoring device and sample process lines, there is no adverse impact on any operating parameters, safety limits or setpoints. Therefore, the margin of safety as defined in the bases to any Technical Speci-fication is not reduced by the addition of this EMF.
Based on the above analysis there are no unreviewed safety questions involved with this modification. Also, inclusion in
-Technical Specifications is not required due to the fact that releases from the Annulus are currently monitored by EMFs required l by Technical Specifications. !
l CN-10879
Description:
This modification will provide proper maintenance and calibration capaoility for the RVLIS transmitters and filled capillaries. The filled capillary systems are connected to the= 4 Reactor Coolant System by sensing lines which connect to the Reactor Vessel head, the seal table, and Hot Legs A & C.(3) I&U requires two isolation valves and a test tee assembly in each of these lines. This modification provides the necessary equipment 3 for the RV bead and seal table lines.
The isolation valves are to be installed in Class B tubing and are ;
QA Condition 1. The structural supports for the Reactor Vessel head sensing line and added valves have been-reviewed by Civil for the effect on the bellows assembly support. The structural ;
supports for the valves and tubing on the seal table connection are typical I&C details. There are no electrical aspects of this ,
NSM. The supports are adequate for all design loads including seismic as determined by the pipe class. No failure modes are created by this NSM. Double isolation is provided between the l test tee and the existing isolation valves (NC279 & V295) as i required by I&E, Safety Evaluation: The valves and test tee being added are of a f consistent class rating (b) and quality of materials and construc- g tion, Therefore, the probability of a previously evaluated 1 accident (in this case, a small LOCA) or malfunction of equipment j-is not increased as a result of this modification. Although this ;
system does not initiate any automatic accident mitigation equip- !
ment, it does provide the operator with information that could lead to manual operator action. The addition of these valves for calibration and maintenance purposes will not af fect the RVLIS system normal or post accident operation. Therefore, the conse- :
quences of previously evaluated accidents or malfunctions of equipment are not increased. No new failure modes are created and i no new accidents or malfunctions of equipment can be identified. 4 No safety limits, setpoints or parameters assumed i~n any safety .
analysis have been affected. Therefore, the margin of safety as J defined in the bases to the Tech Specs has not been reduced. (
CN-10881
Description:
The level of the Unit 1 Boric Acid Tank is not being :
accurately indicated because the tank is not always vented to :
l atmosphere. This modification revises the tank level instrument j j
l-l l
1
7 ,
6 scheme by using a reference leg-type installation which would i indicate properly under pressurized conditions. This modification affects FSAR figure 9.3,4-5_-
'1 Safety Evaluation: This modification will assure that accurate tank levels are being indicated in the control room, on Auxiliary Shutdown Panels, and to Process Control Cabinets, Accurate Boric Acid Tank Level is essential to prevent Tech Spec violations.
This modification will improve the accuracy after Boric Acid Tank Level Transmitters by adding reference legs to account for gas pressure bu11oup beneath the tank diaphragm. This will ensure that the required tank level will be maintained in the Boric Acid _ '
Tank, thus making system operation safer. f CN-10902
Description:
The Auxiliary Feedwater (CA) System assures suffi-cient feedwater supply to the steam generators (S/Gs), in the event of loss of the Condensate /Feedwater System, to remove energy stored in the core and primary coolant. The CA System may also be '
required in some other circumstances such as evacuation of the main control room or cooldown after a loss of coolant accident' for a small break, including maintaining a water level in the S/Gs following such a break. (Ref. 1)
Flow transmitters ICAFT5090, ICAFT5100, ICAFT5110 and ICAFT5120 monitor flow to S/G 1A, 1B, 1C and 10 respectively. These transmitters are overranged during unit startup and plant transients. These transmitters provide indication in the control t room and inputs to the Operator Aid Computer (OAC) and Safety .
Parameter Display System (SPDS). When these transmitters over-range, the SpDS alarms. (Refs. 3, 6) :
This NSM will install additional flow transmitters so that all postulated CA flow rates will be covered. The new transmitters will be connected to OAC inputs as are the existing transmitters.
The SpDS will receive input, as it does now, from the OAC but it will now receive signals originated by the new transmitters. This ,
is a computer software change. The new transmitters will be wired directly to the OAC and will not connect to the existing trans-mit+ers. No other changes will be made to the existing trans-mitters, control room indicators or computer points (Ref. 4). The new transmitters will not be connected to any. safety related
-controls in the control room.
Safety Evaluation: The CA System and the OAC are affected by this NSM. The CA System is QA Condition 1 and the OAC is non-safety.
The operation of the CA System will be unaffected by this NSM.
The new transmitters will be installed in QA1 pipes to safety standards (Ref. 4). The transmitters are to be mounted and supported in accordance with station support criteria and they are uvironmentally qualified (Ref. 5). New cables to be installed are all non-safety and a 10 CFR 50 Appendix R review was per-formed. 'The transmitters will be QA 1, fed by non-safety power and connected to the OAC with non-safety power. These trans-mitters will serve no safety function but are QA1 since they
7 t
serve as part of the CA presture boundary. The SPDS will receive a signal from the OAC as it does now. No new failure modes will be created. . Cabling for the new transmitters will be separate from existing transmitter cables.
This NSM does not involve any FSAR evaluated accident or equipment malfunction initiators and therefore will not increase the probability of such. The additional information available in the control room and the elimination of an invalid SPDS alarm will enhance operator performance and will not increase the conse-quences'of an FSAR evaluated accident or equipment malfunction.
The new transmitters will be installed to safety standards and no ,
new failure modes will be created. Therefore, this-NSM will not l; create the possibility of a different accident or equipment malfunction than evaluated in the FSAR. The new transmitters will )
be used to verify operation of-the CA System, will create a better t working environment, be eliminating an invalid alarm, and will not l affect any key safety parameters. Therefore, this NSM'will not :
reduce the margin of safety as defined in the bases to any Techni- .
cal Specification.
CN-10938_
Description:
These modifications will add loop seals to the ;
condensate drain lines coming cff the control room area and j control room air handling units (CRA-AHU-1 and CR-AHV-1 respec-tively). Presently, there are four (4) condensate drain lines coming off of CRA-AHV-1 and two (2) condensate drain lines coming ,
off of CR-AHU-1. In each case the. condensate. lines are headered i together then go through a loop seal prior to going to the Floor l Drain Sump D. After completion of these modifications, each ;
condensate line will include a loop seal prior to being tied !
together. This modification affects FSAR figure 11.2.2-8.
l Safety Evaluation: These modifications will prevent air from (
being drawn into these QA Condition 1 air handling units (AHU)- a from the condensate drain lines which do not drain properly due to j insufficient back pressure. These modifications will not _ impact i the function of the WL system condensate drain lines or the AHV's.
In addition, the probability of AHU failure due to clogged conden-q sate drain or by any means is not increased. Pipe class and size will not change, t Therefore, the probaM 11ty of an accident or malfunction of -
equipment important to safety as previously evaluated in the FSAR q; will not be increased. Since no new failure modes have been $
created, the possibility of an accident or malfunction of equip- !
ment important to safety which is different than any already l evaluated in the FSAR will not be created. Because the ability of i the AHU's to fulfill their function is not degraded, the conse- :l' quences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. No degradation of safety limits has occurred as a result of this f modification and the reactor vessel core parameters are unaffected q by this modification. The reliability of the equipment associated l with this modification is not degraded. Therefore, the margin of l 1 4
8 safety as defined in the bases to any Technical Specification is
, not.affected. Based on this discussion, no unreviewed safety questions are judged to be involved with this modification.
CN-10968
Description:
The Steam Generator Blowdown Recycle (BB) pumps have had numerous failures duu to higher than design blowdown flow, and blowdown flow flashing and cavitating. These problems can, in part, be attributed to the two phase flow the pumps can see.
A short run of piping with a manual isolation valve will be routed around the pumps. Also piping will be routed from the hotwell pump's discharge to the BB pump bypass line and will include an isolation valve. This will provide subcooling for the BB flow'and prevent flashing,
- c. The interlock closing valve BB39 on low flow is deleted. BB39 will now close on low-low tank level.
~
Safety Evaluation: This modification will not change the function of the system. The changes will result in a more efficient operating system. The NSM will only affect the non-safety related portion of the BB system located in the Turbine Building. The deleted electrical interlock closed BB39 on low flow. This caused ,
the steam generator blowdown pumps to recirculate which prevented -
the pump from taking suction from a tank filled with mostly steam. '
The new interlock accomplishes the same purpose by using low tank level. The proper design conditions are used and overpressure has been considered. Hangers, stress analysis, and ALARA were ,
reviewed. (Ref. 3 and 5)
The modification does not involve or affect any part of a system, including the containment isolation portion of the BB system, associated with any accident discussed in the FSAR. The BB system is noc an initiator for any accidents previousiv evaluated in the
- FSAR or new accidents; therefore the probability of their occur-rence will not increase due to this NSM. The consequences of any accidents will not be increased since this NSM does not affect any
- accident mitigating system or reactor coolant system parameters.
No equipment important to safety is directly affected by these changes. Normal system operation can be attained without the use of the BB pumps, which will decrease system maintenance and increase system efficiency. With overpressure considered, the probability or consequences of a malfunction of equipment impor-tant to safety previously evaluated in the FSAR will not be increased. No new equipment except piping was introduced and t system performance is improved; therefore, the possibility of malfunctions of equipment important to safety different than any already evaluated in the FSAR will not be created. Since no ^;
functional changes are being made to any safety systems and there are no relevant parameters in the Bases to any Technical Specifi-cation, the margin of safety as defined in the Bases to any Technical Specification is not reduced.
9 There are no unreviewed safety questions associated with these NSM's.
CN-11009
Description:
The piping downstream of relief valve 18B28 and 1B8161 is routed to the Turbine Building roof, and may collect .!
rainwater and condensation from valve leakage. The resulting back pressure from this collected water was not ir.corporated in the determination of the relief valve sizing and set pressure. A !
continuous drain to the Turbine Building Sump will eliminate the accumulated back pressure and not affect valve operation. ,
NSM CN-11009/00 and CN-20400/00 will provide a one-inch class G carbon steel drain line downstream of-relief valves 18B28 and IBB161 respectively. The lines will tie in downstream of the 10x16 reducer at 18828 and the 8x14 reducer at 1B0161 (FSAR Figure 10.4.8-10). The' lines will be routed to the Turbine Building trench.
Safety Evaluation: All reference piping is non-safety related and is located.in the non-QA Turbine Building. The discharge points of the relief valves are not monitored release points so the new drain piping does not add any release points requiring monitoring.
In the event a level of activity in excess of a preset limit is detected, the S/G water sample monitor terminates S/G blowdown flow and prevents release of activity to the Turbine Building sump. Due to the large ratio of relief piping to drain piping diameters, too little discharge stream will be diverted to the drain piping to create a personnel hazard.
CF-11104
Description:
This modification will delete the existing pneumatic Low-Low Lube 011 and Overspeed Trip schemes. The three Lube Oil pressure sensors will be replaced with pressure transmitters, current alarms, and electrical tr_ip logic. The Overspeed Trip
- will be-replaced by the Electronic Overspeed Trip Modification available from the vendor to improv+ response time to overspeed conditions and gain independence frotii pneumatic maintenance '
problems. This modification affects FSAR Figures 9.9.7-3 and
- 9.5.4-1.
- Safety Evaluation: The D/Gs are not the initiators of any FSAR evaluated accidents or equipment malfunctions. Therefore, this modification will not increase the probability of any previous 1v i evaluated accidents or equipment malfunctions. This modification
~
will install logic using the same number of independent measure-ments and the same logic schemes. Therefore, this modification will not increase the consequences of an accident or equipment malfunction evaluated in the FSAR. Since no new failure modes will be created, this modification will not create the possibility of a different accident or equipment malfunction than previously evaluated in the FSAR. Aiso, since there are two redundant D/Gs per unit, this modification will not reduce the margin of safety I as defined in the bases to any Technical Specification.
l l
10' ;
i ,
~
CN-11166.
Description:
The nitrogen backup on D/G 1A control air is no longer needed. This is because the " Starting Air" air quality has
-been'shown to be-reliable. This change deletes the nitrogen backup and returns the control air system to its original design.
This change affects FSAR Figure 9.5.6-1.
Safaty Evaluation: This modification is restoring the '1A Diesel's ,
Starting Air System to its original state before the nitrogen backup system was installed. The D/G air quality has signifi-cantly improved with respect to moisture, to an acceptable level, due to maintenance performed on the system components. Since the system will be restored to the previously acceptable configuration as shown on all documents before the nitrogen backup modification took place. This system is safety-related but the change is putting the system back in a state that had already been evaluated as acceptable.
CN-20023
Description:
The Reactor Vessel Level Instrumentation is not j supplied with sufficient cooling air. Design Engineering / Elect.
has-identified overheating problems with Motor Control Centers located-in Rooms 484 and 360 respectively. This modification will provide additior.a1 cooling for rooms 484 and 360. >
Safety Evaluation: The electrical penetration rooms are not supplied with sufficient cooling air. This modification will make the equipment function more reliable because adequate cooling air will be supplied.
CN-20074-
Description:
These modifications will alter the way that EMF-31 gets flushed. Presently EMF-31 is flushed in the same direction that its process fluid flows and the water for this flushing is provided by conventional low pressure survice water (RL). Upon. ,
completion of this modification, EMF-31 will be back flushed with water from the demineralized water (YM).
Safety Evaluation: The functional capabilities of EMF-31 which are to provide radiation information about the Turbine Building.
Sump and based upon that information control the discharge path of the sump will be the same following completion of the modifi-cation. EMF-31 is a non QA condition component and does not '
perform a role in mitigating any design basis accidents.
No plant parameter will be affected by this modification. The YM system will be isolated by a check valve from EMF-31 to prevent contamination of the YM system. Since this modification has no interactica with equipment, structures or components important to safety and doer not create a new release path, there is no in-crease in the probability of accidents of malfunctions of equip-ment important to safety previously evaluated in the FSAR. Since all components modified are not degraded, there is no increase in the consequence of accidents or malfunction of equipment important to safety already evaluated in the FSAR. For both of the above reasons, no possibility is created of accidents or malfunctions of equipment important to safety different from any already evaluated ,
-11 in the FSAR. No plant parameter or setpoints are altered by this NSM so no margin of safety as defined _in the basis to any Tech.
Spec, is reduced.
There is no USQ associated with this NSM.
CN-20090
Description:
This modification will provide for operational enhancements to the Containment Chilled Water System (YV) and ;
YV/RN (Nuclear Service Water) interaction. The purpose of YV system is to supply chilled water to the Lower Containment
, Ventilation' Units, Incore Instrumentation Room Ventilation Units, and the Reactor Coolant Pump Motor Heat Exchangers. The modification will provide annunciation of YV high temperature in the control room. Swapover logic (from YV to RN) will also be modified. Starting circuitry of the-YV chiller will be modified to provide capabilities consistent with the design of the chiller.
Also the removal of a YV trip on closure of (valve) RN437B will occur under this modification. Other sensors currently in use
- sense conditions caused by a valid closure of RN4378 and will trip -<
YV. This modification affects FSAR Figure 9.2.8-7.
Safety Evaluation: The YV system is non QA condition and performs no safety function. The containment isolation function of RN-437 is safety related, but will be unaffected by this modification.
Cooling is not provided to any load by YV during LOCA conditions nor is YV cooling taken credit for in the accident analysis.
Also no system or piece of equipment will be adversely affected by this modification, and the functional capabilities of both YV and RN will be the same following this modification; therefore, the probability or consequence of an accident previously evaluated in the FSAR is not increased nor is the probability or consequences of a malfunction of equipment important to safety previously evaluated in the FSAR increased.
The nature of the YV system and this modification are such that it precludes their potential role as an event initietor. Therefore, the possibility of an unanalyzed accident is not created by the installation of this modification.
Since no piece of equipment will be impacted negatively by this -
modification, no possibility is created of an unanalyzed malfunction of equipment important to safety, u
CN-20156
Description:
-This modification provides vent pipe from relief valves 2CM148, 2CM151, 2CM157, 2CM160, 2CM167, 2CM170, 2CM173, 2CM176, 2CM179 and 2CM182. "D" and "G" feedwater heater relief ,
valves are deleted.
L Safety Evaluation: Low pressure feedwater heater relief valve piping is overstressed during valve actuation. In addition, the valves are vented to the atmosphere in the vicinity of the heaters
l 12 i posing a safety hazard to personnel in the area during a valve opening. This modification will alleviate these problems by:
a) providing additional restraints l b) routing discharge lines from valves close to the floor grating so that relief steam is directed into a funnel 1 drain.
1 Also, relief valves 2CM173, 167, 179 and 182 will be reset to 950 .
psig (ibid); within the 110% of piping design pressure (=880 psia) as required by ASME Code allowables. "G" and "D" feedwater heater relief valves 2CM148, 151, 154, 170 and 176 will be removed from the system. Alternate relief paths exist which make the valves to be deleted unnecessary. These modifications will meet applicable design criteria so will not increase the frequency of unit trips and'are a personnel safety enhancement. All components to be affected are non-QA and are located in the Turbine Building, a non-QA structure. Equipment important to safety is not degraded by this NSM.
CN-20185
Description:
The general intent of this modification is to add ,
the capability to process Ventilation Unit Condensate Drain Tank (VUCDT) contents through disposable vendor demineralizers provided by a contractor. The diaphragm in the recycle monitor tanks will be removed to increast usable tank volume. This modification affects FSAR figures 11.2,2-7 and 11.2.2-11, Sections 11.2.2.1.7, 11.2.2.2.1.10, 11.2.2.2.3.4, 11.2.2.7.1.7- and Table 11.2.2-3.
Safety Evaluation: The VUCDT system was designed for the normal discharge path to be through a radiation monitor to the Turbine Building (TB) sump. If the monitor's high setpoint is reached, >
the discharge stream is automatically isolated from the TB sump and is realigned by an operator to the liquid radwaste release L path. Experience at both McGuire and Catawba shows that higher levels of activity than were anticipat'ed will be present. Use of vendor demineralizers will help expedite the processing of the t
. condensate, which could otherwise require more elaborate processing along the same pathways as the main liquid waste stream, thereby impeding processing of radwaste.
The changes to the-system including additional piping and valves are located in the Auxiliary Building and are designated Duke Class E, QA Condition 2. The piping is hung by construction hangers and is non-safety related. The pipe routing was reviewed for ALARA concerns and for pipe rupture concerns. Fire barrier integrity reviews for piping are adequate.
Electrical instrumentation added to the control panel has no seismic considerations.
The removal of the Recycle Monitor Tank (RMT) diaphragms will not introduce any additional offsite release considerations. However, the venting of these tanks to the room should be reviewed for
13
< ALARA concerns. The probability and consequences of the radwaste
. tank accident is not affected by modifications to the RMT due to its intended function and protection afforded by the Auxiliary Building.
=This system has no emergency or safety function and is not fire protection related.
CN-20201 Descr'iption: This modification will install a radiation monitor (EMF-60) for the purpose of detecting the presence of noble gasses in the Annulus. The monitor is to be located in the Auxiliary-Building (Aux Bldg) and connected to a sample line that draws annulus air samples from-four points within the Annulus. No connections to the Annulus Ventilation (VE) system are required.
This sample line will have a manual valve and solenoid valve at each sample point. Additionally, three solenoid 91ves are located in the Aux Bldg, one each for inlet and outlet isolation and one fcr Aux Bldg air purge of the sample line. The sample return line is routed back to the Annulus.
Safety Evaluation: EMF-60 and the associated piping are non safety related. The components of this sampling system are not required to be seismically qualifted but have been evaluated for potential seismic interaction. The piping associated with this NSM will only penetrate into the Annulus and not into Containment and therefore containment isolation will not be requirea. All radiation released from Annulus will still be filtered by the VE system before release to the environnent.
EMF-60's drawing out of air samples from the Annulus and returning
-the sample back to the Annulus will have no impact on the ability- .
of the VE. system to maintain a negative pressure in the Annulus ;
following an accident. Also, the pipe break failure mode of this -
addition will not prevent the VE system from creating a negative pressure. The reason for the lack of impact is due to the vast differences in the air handling capabilities (VE system 9000 scfm, EMF-60 2-5 scfm).
Since the sample line has the possibility of carrying highly radioactive air from the Annulus to the EMF which is in the Aux Bldg, and automatic isolation / purge feature (activated on sensing high radiation in the sampling air) will isolate the inlet to the EMF and then purge the sample line with Aux Bldg air. The radioactive air will be purged into the Annulus. This feature of the sampling system will avoid concern of high radiation exposure to equipment in the area of the EMF and the Aux Bldg portion of the process line. Health Physics and procedures will control the use of the sampling sy dem in a bypass mode which permits use of the EMF even though a high radiation signal is actuated. These controls are to avoid equipment radiation exposure concerns.
Since no system or piece of equipment is adversely af'ected by the installation of EMF-60, the probability or consequence of an accident previously evaluated in the FSAR is not increased nor is
q:
14
/
the probability or consequence of a malfunction of equipment important to safety previously evaluated in the FSAR increased.
The rature of the EMF and the sampling lines are such that it precludes its potential role as an event initiator. Therefore, the possibility of an unanalyzed accident is not created by the installation of the EMF. I The installation of this NSM will have no adverse impact on any piece of equipment; therefore, it does not create the possibility-of an unanalyzed malfunction of equipment important to safety.
Since this NSM involves only the addition of a monitoring device and sample process lines, there is no adverse impact on any operating parameters, safety limits or setpoints. Therefore, the margin of safety as defined in the bases to any Technical Specif1- ~
cation is not reduced by the addition of this EMF. j 1
CN-20346
Description:
An ATWS is an anticipated operational occurrence (such as loss of feedwater, condenser vacuum, or offsite power) which is accompanied by a failure of the Reactor Trip System (RTS) to shutdown the reactor. The ATWS Rule requires specific improve-ments in the design and operation of commercial nuclear power facilities to reduce the probability of failure to shutdown a o reactor following an anticipated transient and-to mitigate the consequences of an ATWS event.
The ATWS Mitigation System and Actuation Circuitry (AMSAC) that I will be installed at Catawba Nuclear Station is based on the l Westinghouse Owners Group WCAP-10858-P-A, Rev. 1, generic design s
- 3. Design-differences for Catawba Unit 2, which has D5 stream i generators, are specifically identified. The AMSAC design for Catawba is based on conditions that indicate a loss of main -,
feedwater event, wnich if accompanied by a failure of the RTS to l scram leads to overpressurization of the Reactor Coolant (NC)- !
System. The system will monitor the position of all main J feedwater control and isolation valves and the operating status of j both main feedwater pumps.
AMSAC actuation will occur when either both main feedwater pumps trip or when main feedwater flow to the steam generators is ;
blocked due to valves closing in the line. When an actuation occurs, the AMSAC circuitry will perform the following:
- 1) Trip the main turbine
- 2) Start both motor driven auxiliary feedwater pumps
, 3) Close the steam generator blowdown and sampling valves Annunciators, a status indicator and computer alarms in the control room will also be installed.
Safety Evaluation: This modification will affect the Feedwater (CF) and Auxiliary Feedwater (CA) Systems. The CF system takes the treated Condensate System Water, heats it further, and a
)
m u
15 -l r
delivers it at the required flow rate, pressure, and temperature to the steam generators. The CA System serves as a backup for the CF System and is designed as a means to dissipate heat from the NC ,
when normal systems are not available. The FSAR Chapter 15 7 accident in Section 15.2.7 (Loss of Normal Feedwater Flow) is the applicable design basis accident. (Ref 1) This modification will not adversely affect this analysis but Sections 15.2.6, 15.2.7 and possibly others will need to be revised when this NSM is in-
+
stalled. The AMSAC circuits are all non-safety.
Seismic reviews have been completed for mounting the bypass switch on the control board and for mounting the limit switches onto the control and isolation valves. A 10CFR50 Appendix R review has '
also been performed.
A principal criteria' applied to AMSAC is that the.AMSAC functions be accomplished without relying on the existing reactor shutdown system. This modification is designed to separate equipment used ;
for AMSAC and that used for the Reactor Protection System (RPS), i The pressure switches which will monitor the main feedwater pumps will have no RPS interface. The limit switches which will monitor-the main feedwater control and' isolation valves will be used only i by the AMSAC and will provide no signals to the RPS. The AMSAC logic circuitry will have a non-interruptible non-safety 125 V de power source. The Auxiliary Feedwater, Steam Generator Blowdown and Steam Generator Sampling are systems which are safety related or have safety-related components and will receive an AMSAC input.
The interface with these systems is through an existing non-safety / safety isolator and is designed so that the safety related system will perform as designed coincident with a postulated failure of the non-safety AMSAC input.
CN-20363
Description:
The BB pumps have had numerous failures due to.
higher than design blowdown flow, and blowdown flow flashing and cavitating. These problems can, in part, be attributed to the two phate flow the pumps can see. 3 A short run of piping with a manual isolation valve will be routed-around the pumps. Also piping will be routed from the hotwell pump's discharge to the BB pump bypass line and will include an isolation valve. This will provide subcooling for the BB flow and prevent flashing.
The interlock closing valve BB39 -on low flow is deleted. BB39 will now close on low-low tank level.
Safety Evaluation: This modification will not change the function l of the system. The changes will' result in a more efficient operating system. The NSM will only affect the non-safety related I
portion of the BB system located in the Turbine Building. The deleted electrical interlock closed BB39 on low flow. This caused the steam generator blowdown pumps to recirculate which prevented the pump from taking suction from a tank filled with mostly steam.
The new interlock accomplishes the san.e purpose by using low tank l
p 16 i
a level. The proper design conditions are used (Ref. 8) and over-pressure has been considered (Ref. 3). Hangers, stress analysis,
-and ALARA were reviewed. (Ref. 3 and 5)
The modification does not involve or affect any part of a system,
-including the containment isolation portion of the BB system, associated with any accidant discussed in the FSAR. The BB system is not an initiator for any accidents previously evaluated in the FSAR or new accidents; therefore, the probability of their occur-rence will not increase due to this NSM. The consequences of any .I accidents will not be increased since this NSM does not affect any '
accident mitigating system or reactor coolant system parameters.
No equipment important to safety is directly affected by these changes. Normal- system operation can be attained without the use of the BB pumps, which will decrease system maintenance and-increase system efficiency. With overpressure considered, the t probability or consequences of a malfunction of equipment impor- I tant to safety previously evaluated in the FSAR will not be increased. No new equipment except piping was introduced and system performance is improved; therefore, the possibility of malfunctions of equipment important to safety different than any s already evaluated in the FSAR will not be created. Since no functional changes are being made to any safety systems and there are no relevant parameters in the Bases to any Technical Specifi- 1 cation, the margin of safety as defined in the Bases to any Technical. Specification is not reduced.
There are no unreviewed safety questions associated with these NSM's.
CN-20399
Description:
Dams in the Refueling Water (FW) Trench prevent water; which collects between the Reactor Makeup Water Storage Tank (RMWST) and the Refueling Water Storage Tank (FWST), from draining.
Safety Evaluation: The FWST trench contains water barriers or
" plugs" to retain the contents of the FWST in the surrounding missile wall following postulated tornado missile damage to the tank. During rainstorms or extended periods of rain, rain water is retained in the section of trench between one of these barriers and the Reactor Makeup Water Storage Tank. The trench was origi- .
nally designed to provide drainage for this area but this design <
was defeated by the addition of the referenced water barriers.
This modification will allow sections of the trench on each side ;
of the water barriers to communicate for drainage by providing a single pipe through both barriers. The pipe penetrations will be sealed so that the FWST contents cannot leak past the penetration.
Since the only structures affected are the water barriers, and the barriers' ability to perform their safety function (retaining the contents of the FWST) is not diminished, the probability of an accident or malfunction of equipment important to safety previous-ly evaluated in the FSAR will not be increased. By the same reasoning, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR
~
k 17 ,
P P is not increased. The added drain pipe provides a flow path for-water from the RMWST if that tank should rupture and spill-into the trench. This accident is not initiated by the NSM's and no new accident or malfunction of equipment important to safety not
- already evaluated in the FSAR is created. The fact that spilled _ ,
water could now enter the FWST trench does not create a new consequence more severe than those already evaluated in Chapter 15 of the FSAR. No margin of safety defined in the bases to any Tech Spec is reduced, There are no Unreviewed Safety Questions associated with this NSM.
CN-20428
Description:
In order to facilitate the drainage of Containment Penetration M-235 for Type C leak rate testing, a drain valve will' be innta11ed upstream of the NM6A bypass line discharge, r
Safety Evaluation: The proposed modification involves the Nuclear Sampling (NM) system, which provides samples from various nuclear related systems _for chemical and radiochemical analysis. The >
added drain valve will be located on the pressurizer steam space sample line. This line is a part of the reactor coolant pressure boundary (RCPB), but due to its small size (1/2") it is Class B
~
not A. The drain line which is to be installed will also be Class B piping.
The major. safety function of the QA-1 portion of the NM system is to provide containment isolation on appropriate safety signal.
The system also serves to maintain the RCPB integrity. Neither of these functions is impaired by the addition of drain valve. A-seismic analysis has been performed to assure the seismic qualification'of the modified sample line.
The failure of an instrument line connected to the reactor coolant system is postulated as part of the FSAR accident analysis, The addition of a QA-1 drain line to the instrument line does not make this accident more probable.
This modification does not adversely affect any plant safety functions, so the consequences of this or any other accidents are not increased.
All potential instrument line failures are clearly bounded by-the rupture of a 3" CVCS line. In-addition, no new failure modes or operating characteristics are created by this modification.
Therefore, no new accident scenarios are created.
Since no new failure modes or operating characteristics are introduced, the probability of previously evaluated malfunctions of equipment important to safety are not increased, and the possibility of new malfunctions is not created.
No plant parameter that is either controlled by Technical Specif1-cations or that has any adverse impact on the safety / design
t 18 analyses is affected by the modification. Thus, no margin of safety is reduced.
Therefore, based on the above analysis; no unreviewed safety questions are involved with this NSM. >
CN-20486
Description:
This modification will replace the diesel generator.
(D/G) safety-related pneumatic engine shutdown logic. The existing pneumatic logic for Lo-Lo Lube Oil and Overspeed trips The existing Lo-Lo-Lube will be replaced with electrical logic.
011 Pressure sensors will be removed and the pneumatic tubing will
, be isolated. Three pressure transmitters will be installed for the 2-out-of-3 trip scheme and the electrical trip logic will be designed using existing control circuits. The transmitters will be mounted on a seismically qualified' instrument stand near the instrument location for D/Gs IA and IB and on existing seismic pipe hangers near the instrument locations for D/Gs 2A and 28.
The Overspeed logic will consist of two independent magnetic pick-ups connected to speed relays to be used'in a 2-out-of-2 trip "
scheme. This is a vendor supplied package. The new relays and instruments will be placed in existing D/G control cabinets and in '
.a new QA1, seismically mounted, cabinet. The new control cabinet will be installed next to the existing control cabinets in the D/G rooms. New cables will be pulled within each D/G room between the D/G and the cabinets. No cables will be pulled outside the D/G rooms.
Safety Evaluation: The D/Gs, control cabinets and the Lo-Lo Lube Oil Pressure and Overspeed trip systems will be affected by this modification. All camponent and systems affected by this modifi-cation are-QA Condition 1. The D/Gs are designed to provide standby AC power to equipment required to safely shutdown the reactors in the event of a loss of p wer to-the plant emeroency buses. The appliccble design basis accident is described in FSAR Section 15.2.6 (loss of Non-Emergency AC Power to the Station
-Auxiliaries). .
A seismic analysis has been performed for the mounting of all instruments, relays and cabinets an all the new equipment is qualified for the D/G room environment. No cables will be pulled outside the-D/G rooms so there are no 10CFR50 Appendix R concerns.
The electric trip logic will not change the function of the j systems involved. The engine overspeed trip setpoint will' remain the same and two independent speed measurements will be used in a 2-out-of-2 logic scheme; the same as exists now. Three indepen-dent oil pressure measurements will be used in a 2-out-of-3 logic scheme for the lube oil pressure trip; the same as exists now.
The system logic and details of the system design are discussed in the Final Scope Document along with details about the failure modes. No new failure modes will be added by this modification.
19 FSAR Section 8.3.1.1.3.4 (Diesel Generator Protection Systems) mentions "an overspeed control pneumatic. trip" and therefore will need to be revised for this modification. No Technical Specification changes are required. 1 CN-20535
Description:
This modification will move the Unit 2 (Model 0-5) steam generators' narrow range lower tap- frorn its present location above the transition cone to an existing (plugged) tap located ;
below the transition cone. Impulse lines will be rerouted as necessary. New narrow range level transmitters will replace the existing transmitters. Temporary level instrumentation installed per NSM CN-20303 will be removed. Steam generator normal ;
operating. level and alarm setpoints will be adjusted, but no '
. protection / control. function will change.
Safety Evaluation: This steam generator water level instrumenta-tion is safety-related. It is a part of the Reactor Trip System (RTS) and the Engineered Safety Features Actuation System (ESFAS).
The RTS trips the reactor on low-low steam generator water level.
The ESFAS actuates the Auxiliary Feedwater pumps on low-low steam generator water level and actuates feedwater isolation and turbine trip on high-high steam generator water level._ 'This level instru-
, mentation also provides input to the Main Feedwater (CF) control- ,
system. '
For the RTS and ESFAS there are four separate channels per steam generator and the safety function is actuated on two out of four logic from any steam generator. The low-low level trip protects the rea:: tor from loss of heat sink and high-high level trip -
mitigates steam generator overfill transients and protects the steam lines and turbine from water-induced transients.
New level trip setpoints have been evaluated by Westinghouse and MNSA for the ESFAS and RTS safety functions. These new setpoints consider the new instrumentation range, inaccuracies, time re-sponse, etc. The following FSAR Chapter 15 accidents have been-
, preliminarily evaluated assuming the new set points:
Loss of External Load / Turbine Trip (FSAR Sections 15.2.2 &
15.2.3)
Steam System Piping Failure (FSAR Section 15.1.5)
Loss of Non-Emergency AC Power to the Station Auxiliaries (FSAR Section 15.2.6) l -
Loss of Normal Feedwater Flow (FSAR Section 15.2.7) <
Feedwater System Pipe Break (FSAR Section 15.2.7)
Feedwater System Malfunction Causing an Increase in Feedwater
- j. Flow (FSAR Section 15.1.2) t
=
m
i p I20 i;.
Mass and Energy Release Analysis for Postulated i Secondary System Pipe Ruptures Inside Containment (FSAR Section 6.2.1.4)
The following accidents were re-analyzed because they are the accidents most sensitive to a reduction in the low-low level trip setpoints:
Low of Non-Emergency AC Power to the Station Auxiliaries (FSARSection15.2.6)
Loss of Normal Feedwater Flow (FSAR Section 15.2.7)
Feedwater System Pipe Break (FSAR Section 15.2.7)
The evaluation of the above accidents with the new level setpoints concludes that the' design basis functional requirements are still satisfied. Technical Specification changes are being submitted to ,
the NRC for approval concerning new level trip setpoints in Tables 2.2-1 and 3.3-4. Changes will be made to Chapter 15 of the FSAR to reflect the new Westinghouse safety analysis. O The narrow range level instrumentation serves'a dual protection-control function. The failure modes and effects analysis is still-valid due to the two out of four trip logic since the control channel could' fail low or high and the RTS or ESFAS safety func-tion would still prevail due to the valid remaining two out of three trip logic.
The present transmitters are being replaced with new environ-mentally qualified transmitters in the same location with the same seismic mounts. Cable routing to the transmitter is unaffected, therefore Appendix R criteria does not have to be re-evaluated.
Stress analysis and piping interactions of the impulse lines have been considered. Westinghouse determined that the instrumentation accuracies were adequate without insulation covering the new segments of impulse line. Since existing (plugged) taps are being j used by the new instrumentation and the old lower taps will be welded closed per QA1 pipe specifications, the integrity of the steam generator pressure vessel is being maintained. <
CN-20539
Description:
This modification is being implemented as a result of the findings in NRC IEB 85-03 "Mov Common Mode Failures During Plant Transients Due to improper Switch Settings".
Selected valve operators in the NC, NS and ND systems (also BB system on unit 2 only) are being modified to resolve some concerns L for these valves not attaining their desired positions because of insufficient torque switch settings. relative to the resistance through the stroke travel. The specific valves being modified are ;
discussed in the NSM package referenced in this calculation.
I I'
l
B 21 The subject valve operators have " limit switch" actuated " torque bypass switch" contacts. When the limit setting is reached, the torque switch will trip off the motor when unusually high torque is measured signifying some valve abnormality. Presently, most of these valves engage the torque switch at a small value of open stroke position 5%. This-sometimes causes the motor to trip off when the valve is not yet fully unseated, still partially closed.
To assure that these vales open fully, the " torque bypass switch" will be set to defeat the torque switch for a much longer opening -
stroke travel 0 to 50% i 25% open. In this manner, the maximum motor. torque will unseat and open the valve against full flow and a p without allowing the torque switch to operate until the valve is at least 25% open. Then, if for some reason an unusually.high 1 torque is measured, the torque switch could trip off the motor, protecting the valve but not leaving it closed when it should be i open. The torque switch protects the valve when opening by tripping the motor when the back seat is reached due to increased torque (resistance), r The above discussion is for the motor open (M/0) circuit. The motor close (M/C) circuit will be modified in a similar manner for the same reasons on selected valves. The torque switch will-still determine the seating forces and control " stopping the motor" in the close stroke.
Safety Evaluation: Functionally, these valves will operate identically to the way they presently operate. With the new control circuit wiring and " torque bypass switch" setting, the valves should be more reliable in attaining the desired positions, both open and closed, without letting the torque switch protection device interfere. Also,'other indications'and interlocks associ-ated with valve operation, position etc., will not be affected by these changes. No new failure modes will be created as a result of this modification. Appendix R and environmental qualification have been appropriately considered and there are no concerns in these areas. No power supplies or breakers are affected.
The torque bypass contacts are being adjusted to a travel span of 50% i 25%. Valve torque bypass travel spans are not relevant to the initiation of accidents. Therefore, the probability of such-is not increased. .These valves will operate (open, close) identi-cally to the way they presently operate but should be more reli-able in attaining their desired rositions when required to change. '
Therefore, the consequences of accidents and malfunctions of equipment are not increased. No new failure modes are created as a result of the torque bypass travel span increase and a premature motor trip resulting from high torque is removed. Therefore, the ,
probability of a malfunction of equipment important to safetv is not increased and no new malfunctior.s are introduced. No new accidents are created based on a logical review of the modifica-tion end its effect on systems operation. No safety limits, setpoints or assumptions in any safety analysis are affected
4 4
22 by this modification. Therefore, the margin of safety defined in the bases to the Tech. Specs is not reduced.
CN-20545
Description:
A Design Study was initiated to monitor feedwater and steam system parameters in an' effort to determine the cause of
-feedwater flow and steam generator level control problems. The recommended action'resulting from this study was to-redesign the main feedwater flow restricting orifices. Westinghouse Eltetric Corporation has provided the design expertise for this modi 11- '
cation. The redesigned orifices will be less sensitive to fouling while maintaining the required pressure drop for the proper flow ,
split to the steam generators.~ Replacement of the orifices should result in improved unit power production since orifice fouling results in use of feedwater bypass control valves, further opening of feedwater control valves and eventual unit power reduction due to control valve limitations.
Safety Evaluation: The Feedwater System (CF) supplies feedwater to the four steam generators at the temperature, pressure, and flow required to maintain proper steam generator water levels commensurate with reactor power output and turbine steam requirements. In response to steam generator tube wear problems experienced by other nuclear stations with steam generators identical to Catawba Unit 2 steam generators, a split feedwater flow arrangement is used to limit flowrate into the main feedwater nozzle at high loads. Feedwater flowrate into the main feedwater ,
nozzles is limited at high loads by diverting a portion of the flow to the auxiliary feedwater nozzles through the feedwater bypass piping. This NSM modifies the flow restricting orifice installed in the main feedline downstream of the feedwater bypass piping takeoff to provide the necessary pressure di.fferential to attain the desired bypass line flowrate.
The orifices located in the doghouse are class F. The pressure drops across the orifice are the same as before. Structural mechanics, stress analysis and hanger design have been reviewed.
(Ref. 3 & 6) The new material is compatible with the system. The orifices function as before.
The probability and consequences of accidents previously evaluated in the FSAR will not'be increased. The orifices are properly designed and function exactly as before. This passive equipment should have no effect different than before on system operation during normal operating conditions or during an accident scenario.
Since the orifices and system will perform as before, no possibi-lity of an accident has been created which is dif ferent than any already evaluated in the FSAR. Again, since system operating parameters have not changed with the replacement of this passive orifice, the probability and consequences of a malfunction of I l equipment important to safety previously evaluated in the FSAR >
will not be increased. The integrity of the orifices has been reviewed by Westinghouse in conjunction with Duke Power and no other new equipment has been introduced to the system. Based on the above, and system operating parameters not changing, possible l
23 malfunctions of equipment important to safety different than any already evaluated in the FSAR will not be created. No safety parameters or design limits have been adversely affected, and no margin of safety as defined in the bases to any Technical Specification is reduced. There are no unreviewed safety questions associated with this NSM. j CN-20550
Description:
The Process Radiation Monitoring System (EMF) l monitors radioactive releases from the station or compliance with 'l the station Technical Specification limits. It also monitors {'
various equipment and systems for abnormal radiological conditions which are indicative of equipment malfunctions or which could constitute an uncontrolled release path. One of these monitors is
-located in.the condenser air ejector exhaust. In the event of a steam generator tube rupture or other primary _to secondary leak, 3 the radioactivity would eventually appear in the condenser and !
subsequently be discharged through the ejector. The condenser air !
ejector monitor continuously monitors the activity of these gases and provides readouts and an annunciator alarm from high levels.
j H
An alarm is also provided for loss of sample flow through the .j monitor which is detected by a vacuum switch inside the monitor. i This switch is subject to corrosion and possible failure if the !
sample flow contains water vapor and to prevent this, an air 1 filter which acts as a moisture separator is installed in the flow line. This filter, however, has proven ineffective in removing a sufficient amount of water vapor to prevent switch degradation. !
This NSM will replace the existing air filter with a refrigerated compressed air dryer. The dryer will be mounted on the EMF skid where the air filter is now located. Electrical power will be provided from the 600 Volt Blackout Auxiliary Power System which '
supplies the monitor skid. The existing drain line from the filter will be replaced with a larger drain line for the dryer and a bypass line with valves will be installed to provide a flowpath around the dryer for maintenance purposes. A drain connection will also be added to the discharge flowpath piping of EMF 33 to ,
facilitate removing any collected moisture in the monitor should I the dryer fail or be removed from service for maintenance. j Safety Evaluation: The air filter to be replaced, associated !
piping, EMF 33 which they serve, and the EMF discharge piping have ;
no QA condition assigned and do not perform a safety function. i The electrical power for the dryer is non-1E. The EMF skid is f non-seismic and the dryer will not be seismically mounted. The
. leakoff line from the dryer will drain to the turbine building i sump as it presently does and any radioactive effluent discharged through this path would be detected by the turbine building sump monitor, EMF 31. A failvre of the dryer would not prevent operation of EMF 33 since the bypass line could be opened to "
provide the needed flowpath. Tech Spec 3/4.33 which addresses operability of the monitor will not be adversely affected. -
i-The condenser air ejector monitor cannot initiate any of the accidents described in the FSAR, therefore, the probability of f
~
c 24
,. 1 such accidents will not be increased. The exhaust from the ejactor is discharged through the unit vent which has radiation mo.,vtors also, therefore, no accident of a different type will be created. Since the air dryer and radiation monitor have-no operating characteristics or failure modes which could adversely I affect any equipment important to safety, the probability of a malfunction of such equipment, either previously-evaluated in the -
FSAR or different than any already evaluated, will not be increased. Since all radiation release paths associated with the dryer and monitor are monitored again through other radiation monitors before release to the environment, there will be no-increases in doses to the public as a result of this change, therefore, the consequences of an accident or malfunction of equipment previously evaluated in the FSAR will not be increased.
There are no applicable margins of safety defined for any of the equipment affected by this change, so no margins of safety as defined in the bases of any technical specifications will be reduced.
CN-20555
Description:
This modification installs a. drain for the 2A D/G flow path to-support outage maintenance activities.
Safety Review and USQ Evaluation: The Nuclear Service Water System (RN) provides essential auxiliary support functions to Engineered Safety Features of the station. In conjunction with the Ultimate Heat Sink, comprised of Lake Wylie and the Standby Nuclear Service Water Pond, the RN system is designed to supply cooling water to various heat loads in both the safety and non-safety portions of each unit. Provisions are made to ensure a continuous flow of cooling water to those systems and components necessary for plant safety during normal operation and under accident conditions. Sufficient redundancy of piping and components is provided to ensure that cooling is maintained to essential loads at all times.
The D/G engine starting air compressor af tercooler is supplied constantly as the compressor operates periodically to maintain starting-air tank pressure. Flow is set by a manual throttling valve. Cooling water is supplied to the 0/G engine jacket water cooler only when the diesel is in operation. This is accomplished by an EMO valve interlocked to open when the diesel starts and close when the diesel stops.
If a unit's diesel is out of service or down for maintenance, then the shared valves normally powered from that channel are provided '
with manual switchover to the other unit's diesel of corresponding channel. Therefore, any one diesel generator can be down and the RN system can still shut the plant down safely assuming a LOCA, station blackout, and single failure.
The accidents discussed in the FSAR are not initiated by any systems affected by this NSM, so the probability of an accident previously evaluated in the FSAR will not be increased. The design of the NSM using system design parameters, Class C piping,
p, i 25 Class C isolation valve normally closed, and flanged end piping will not_ functionally affect any system during any mode of opera-tion. Therefore, the consequence of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Likewise, the probability of a malfunction of equipment important to safety previously evaluated in the FSAR-is not increased. Based on the above and that no new common mode failures have been identified and no system functions are affec-ted. Since design will prohibit excessive leakage or potential line break, the possibility of an accident or malfunction of equipment important to safety different than any already evaluated in the FSAR will not_be created. Since no safety parameters or design limits have been adversely affected, no margin of safety as defined in the bases to any Technical Specification is reduced.
There are no unreviewed safety questions ast>ciated with this NSM.
CN-40002.
Description:
Route cable and supply power to the newly installed or relocated equipment in the contaminated trash, clean trash and decontamination processing areas. Relocate telephones and lighting as necessary for the revised layout. Supply power to HVAC controls, i
Safety Evaluation: No unreviewed safety question-is created as a result of this modification. This modification concerns changes to the Auxiliary Service Building facilities for laundry, waste compaction, and decontamination processes. Upgrading equipment-
.will improve operating efficiency, reduce dependency on vendors and reduce operating costs. No safety. system will be degraded as ;
a result of this modification.
CN-40010
Description:
Remove the temporary Admin Building, Terminate all e utility services cap piping and show location on the drawings.
Safety Evaluation: This modification is non-QA and is outside the protected area. This modification is not a part of a system or component that is required for safe operation or shutdown of the-plant. This modification has no affect on nuclear safety.
CN-50070
Description:
Replace existing refractometer units on the Boron Recycle and the Liquid Waste Recyc 4 System with Dynatrol units.
Safety Evaluation: Existing refractometers do not measure concentrates properly for the concentrations that would be seen in operation. This modification should improve reliability and performance of the evaporators and will not adversely impact any safety _ or non-safety system.
CN-50072
Description:
This modification provides sight glasses for the Boron Recycle and the Liquid Waste Recycle Evaporators.
Safety Evaluation: This modification will enhance instrument i reliability and will allow the instruments to be allowed in
26 I testing. This modification does not have any adverse impact on any safety or non-safety system.
CN-50192
Description:
A chemiral feeder unit will be added for chemical i addition to the Radwas,;e Batching Tank, Evaporator concentrates i holdup tanks, and evaporator concentrates batch tanks. Piping will be routed to each tank from the chemical feeder unit water '
and a power supply will be needed for the feeder unit. 7 Safety Evaluation: This modification will add a chemical feeder unit in the wa'ste shipping area. This feeder unit will be connected by piping to the radwaste batching tank, evaporator concentrates holdup tank, and the evaporator concentrates batch ,
tank. This feeder unit will mix and pump chemicals to the tanks. ,
The unit will be located in the waste shipping area to reduce :
personnel exposure and provide easier access. The electrical portion of this modification provides 460 volt a.c. power for the pump and mixer, and modifies the ! & C list to show a sight glass which is installed on the pump. All electrical work is non-QA. !
The addition of this equipment will not affect the ability of the radwaste system to perform its desired function. The addition of this equioment is an operational enhancement. Equipment important to the safety of the plant during design basis events will not be affected by this modification. Since no mitigating system is affected, the margin of safety as defined by the Technical Specif1-cations will not be affected.
CN-50292
Description:
This modification provides the plant tie-ins for the .
Monitor Tank Building.
~
Safety Evaluation: None of the systems to which tic ins will be made are safety-related, nor do they contain any high energy lines. Stress analysis and DE-designed supports are not required i so the new piping will be field supported. However, radiation shielding for the piping passing through the electrical penetra-tion areas and some electrical panels associated with this modifi-cation will be seismically mounted to prevent interactions with safety-related equipment.
Fire protection will be maintained throughout the installation of this modification and any fire barriers that are breached by the installation of wall penetrations will be temporarily plugged and will be restored to operability by appropriate fire seals as the modification is completed. The fire protection system itself is not affected by this modification.
Core drilling through the exterior Auxiliary Building wall will be performed with the unit is cold shutdown and will not affect any equipment required to maintain the safe shutdown condition. The use of the temporary fire barrier plugs described above will also serve to prevent any significant flooding in the electrical penetration room from outside the building.
l
27 No accidents or malfunctions of equipment important to safety are either created by or otherwise affected by this modification. No margin of safety as defined in the basis to any Technical Specification is reduced by this modification and no Tecnnical Specification will require revision. Therefore, no unreviewed safety questions are either created by or involved with this modification.
CN-50330
Description:
This modification will upgrade 1 out of the 2 non-safety level transmitters per pit to safety grade. This will provide a third level transmitter per pit to accommodate a 2 out of 3 logic instead of the present 1 out of 2 logic.
Safety Evaluation: Past experience has shown that a single spurious failure to the " low" position can initiate a "swapover" when there is an adequate water level in the RN pits. Inadver-tently challenging the system with numerous valves changing position and starting all RN pumps is unnecessary and reduces the reliability of the system. The 2 out of 3 logic will eliminate system re-alignment due to single failures except power supply failures. Loss of a particular 120 VAC power supply in each pit would cause 2 of 3 Level Transmitters to read low and initiate swapover. In this scenario the particular power supply is the same one that powers the transfer logic for that pit and swapover is initiated anyway. Thereforc, no new failures that initiate swapover have been introduced. The upgraded level transmitters will still monitor differential level across the screen by compar-ing their level reading to that of the remaining non-safety level transmitter in each pit. 'his is a non-safety function and the safety and non-safety circuits are isolated by an analog optical isolation device to protect the safety circuit.
The existing level transmitters have 120 VAC normal power backed by batteries and Class 1E Diesel Generator Power consistent with the pit they are in. The battery back-up ensures that swapover does not occur upon loss of normal AC power before the diesels t, tart if they are going to start. Also pit level indications (for existing safety grade level transmitters only) will be maintained in the control room independent of normal AC power or the diesel generator status.
The upgraded level transmitters will be QA-1. They will have Unit 1 Class IE diesel generator power consistent with the pit they are in. The two upgraded level transmitters have battery backed normal AC power as described earlier. An Appendix R review has been performed. The new level transmitters will have a low icvel alarm on the computer with no annunciator and a low-low level computer alarm and annunciator. The two existing level trans-mitters have a computer alarm and an annunciator for low level ard lew-low level. All alarms are on a 1 out of 3 logic which pro-vides early warning of a failure or valid low level signal. The failure mode of all the safety grade level transmitters is the same. They fail low on loss of power. This is desirable because ,
the SNSWP is the qualified source for the ultiamte heat sink.
28~
The new arrangement has been reviewed for reliability concerns and 1 judged to have no adverse impact on plant shfety. ,
1 This modification presents no safety concerns and is considered to improve overall system reliability by precluding unnecessary system re-alignments while ensuring swapover on valid loss-of lake <
scenarios. No new failure modes of safety related equipment are !
created. The acciaent mitigation functions of the RN system are not degraded. The Technical Specifications need to be updated to i reflect these changes along with the FSAR. Those affected sections are identified below:
CN-50386
Description:
This NSM will install safety-related simplex ;
strainers in parallel with the existing RN pump lube injection strainers (RN PLI strainers). The current RN PLI strainers are ;
high maintenance components that impact the operation of both units. Raw water from Lake Wylie and the SNSWP regularly clog (
these strainers, requiring them to be disassembled and manually .
cleaned. The design of the RN PLI strainers makes disassembly and !
cleaning a tedious and time-consuming process. Obtaining spare parts has been a problem. Since there are only two RN PLI strainers to support the station (one for each channel), when one strainer comes down for maintenance both units are entered into an '
action statement. If an RN PLI strainer was to become inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, both units would have to shut down.
This NSM will ease the above situation by installing two safety-related simplex strainers in parallel with each existing RN PLI strainer. After this NSM is implemented, if one or both RN PLI strainers becomes inoperable, both units could run indafi-nitely with the safety-related simplex strainers. Only one simplex strainer will be required for a train to be operable. If one simplex strainer becomes clogged, the other simplex strainer can be used while the first one it being cleaned. The simplex strainers are not intended to replace the existing RN PLI strainers, only to provide safety-related backup to the RN PLI ,
strainers.
Each parallel straining path will have inlet and outlet isolation valves. A drain valve will be installed downstream of the strainers to facilitate flushing. The taps for differential pressure measurement will be relocated to facilitate measurement >
across any strainer and alarm the control room on high delta P. -
The old taps will be used as pressure test points in their current location.
The simplex strainers are seismically designed and attached to the floor. New piping is Class C except for some small drains which -
are Class G downstream of a Class C isolation valve. The design pressures and temperatures correspond to system design parameters.
No dose assessment is required. Interaction analysis has been performed. Strainer mesh size meets requirements. Stress analysis is complete. Flooding in the pumphouse has been considered.
l 29 L
i j
Safety Evaluation: The Nuclear Service Water System (RN) provides !
essential auxiliary support functions to Engineered Safety u features of the station. In conjunction with the Ultimate Heat .
Sink, comprised of Lake Wylie and the Standby Nuclear Service ,
Water Pond, the RN System is designed to supply coolitig water to !
various heat loads in both the safety and non-safety portions of
! each unit. Provisions are made to ensure a continuous flow of t cooling water to those systems and compenents necessary for plant safety during normal operation and under sccident conditions.
Sufficient redundancy of piping and compot. ants is provided to ensure that cooling is maintained to essent'ai loads at all times.
These new simplex strainers are an enhancemen to the system and ,
provide backup strainer capability. The PLI ,, vainers can not ;
initiate any FSAR accidents. The new strainer > tre designed for RN pump requirements and seismic loads. Stress p alysis and interaction analysis have been completed. The abtve considera-tions will preclude possible damage to other systems; thus, probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased.
Likewise, the possibility of an accident different than any already evaluated in the FSAR will not be created. The increased ,
redundancy of the strainers will increase the reliability and not hinder performance of the RN pumps which are relied on during cccident scenarios, so the consequences of an accident or malfunc-tion of equipment important to safety previously evaluated in the FSAR are not increased. The RN pumps are more reliable and the new strainers serve as backups and have no debilitating effect on the present strainer or RN pumps; therefore, there is no possibi-lity of creation of malfunctions of equipment important to safety different than any already evaluated in the FSAR. Since no safety parameters or design limits have been adversely affected, no margin of safety as defined in the bases to any Technical Specifi-cation is reduced. There are no unreviewed safety questions ,
associated with this NSM, CN-50391
Description:
The RN pump motor coolers and RN pump motor upper bearing oil coolers are postulated to rob flow from the RN pump lube injection flows during periodic one pump RN flow balances. '
Currently, all flows to RN pump auxiliaries--cooling flows as well <
as lube injection flows- pass through the RN pump lube injection strainers. Since RN pump auxiliary cooling flows do not need to l
pass through the RN pump lube injection strainers, this NSM will move the RN pump auxiliary cooling flows from the RN pump lube injection headers to the RN supply headers downstream of the main RN strainers.
Safety Evaluation: The RN System is a QA Condition 1 system which provides cooling water to both safety and non-safety related '
loads. The piping installed for this modification is Class C and has been seismically qualified. The existing RN strainer is of L the appropriate size for the heat exchangers being supplied RN cooling (RN pump motor coolers and RN pump motor upper bearing oil l
coolers); therefore, the removal of these cooling loads from the
N ' '
30 RN pump lube injection strainer does not adversely affect the heat exchangers (Ref. 3).
I No piece of equipment or system is adversely affected by this modification; therefore, the probabilities or consequence of an accident, previously evaluated in the FSAR, is not increased nor is the probability or consequence of a malfunction of equipment important to safety previously evaluated in the FSAR increased.
The nature of the modification and its impact on the RN System precludes its potential role as an event initiator. Therefore, the possibility of an unanalyzed accident is not created by the installation of this modification.
Since no piece of equipment is adversely affected, no possibility is created of an unanalyzed malfunction of equipment important to safety.
There is no affect to any operating parameter, safety limit or setpoint; therefore, the margin of safety as defined in the bases of any Tech Spec, is not reduced by this modification.
l
n .
St Catawba Nuclear Station Summary of Exempt Variation Notices Completed Under 10CFR E0,59 i
CE-0880
Description:
This change will convert the Unit / Main Turbine Control Valve operation to 3 admission partial are mode. This will require a 3/16" Orifice Y-Strainer to be installed in the ,
drain line before ISM 25 to prevent flashing across the valve. '
This change affects FSAR Figure 10.3.2-4.
Safety Evaluation: Exempt change CE-0880 will install a 3/16" Orifice Y-Strainer in the line up-stream of ISM 25 to prevent flashing across the valve. This is required as a result of revising the Main Turbine Control Valve operation to 3 admission j partial arc mode. All work performed per this Exempt Change is non-safety. Safety-related systems, structurer, or components will not be' degraded by installing this Y-strainer. No changes to the Tech. Specs, will be caused by this change. This modifica-tion will not cause any unreviewed safety questions.
CE-1019 Descriptien: This change removes check valve internals from IFD25 and IFD65. This change affects FSAR Figures 9.5.4-1 and 9.5.4-2. .
Safety Evaluation: This change will remove the internals from check valves IF025 and IF065. These valves currently restrict the NPSH to the Fuel Oil Pump. Since the level requirements set forth i in the Technical Specifications always maintain a flooded pump !
suction from the Fuel Oil Day Tank, these valves are not required. '
1 Due to the fact that these valves are unnecessary and pose i potential NPSH problems, the margins of safety will be enhanced !
rather than degraded. The likelihood or consequence of an i accident will not be created by this modification. A more reliable situation will be attained by removing these internals, thereby increasing the margin of safety. Technical Specifications i will not be affected. No unreviewed safety questions are created i by this modification. ;
I
Description:
I and 2 NBPG 6800, which measure Reactor Makeup
~
CE-1231 CE-1232 Storage Water Tank pressure, were not installed per the appro-l priate instrurnent detail. These changes reinstall 1 and 2 NBPG j 6800 per the appropriate instrument detail. This change affects FSAR Figures 9.3.5-7 and 9.3.5-8. l Safety Evaluation: The reinstallation of 1 and 2 NBPG 6800 does not increase the probability or consequences of an accident Drevio'usly evaluated in the FSAR nor does it increase the pro- !
bability or consequences of a malfunction of safety related equipment as previously evaluated in the FSAR. This change does not create any new accident scenarios or any new possibilities of safety related equipment malfunction different than those evaluated in the FSAR. This change will not reduce the margin of I
I u i
2 ;
safety as defined in any Technical Specification bases. This change is not significant but does affect Figures 9.3.5-7 and 9.3.5-8 (NB System flow diagram) of the FSAR.
CE-1260
Description:
This change replaces existing seals with Five Star !
seals and changes teflon bushings to carbon steel bushings on the ;
Unit 2 'C' Heater Drain Tank Pumps. This change requires that the drain lines off the 'C' Heater Drain tank pumps be modified and '
that 2HWFE-6600 be repiped to allow proper flow indication. This modification will require a change to FSAR Figure 10.4.10-1.
Safety Evaluation:. The replacement of the existing seals with Five-Star seals and the teflon bushings with carbon stee1~ bushings on the 'C' Heater Drain Tank Pumps does not create any unreviewed safety question as evaluated in the FSAR. No safety system will l be degraded, and no functional changes to any system will take .
place as a result of these changes. These new seals and bushings will not affect the integrity of the CF system. The new seals are t; expected to last longer and give more reliable service. The new bushings will stay in place better than the teflon Lushings.
The rerouting of the lines for 2HWFE-6600 is required to have the flow meter function as designed. .
CE-1509
Description:
This change will relocate IRNTE5000 and 1RNTE5010 upstream of Tee branching to the NS Heat Exchanger and KC Heat Exchanger to allow daily monitoring of the RN pump discharge -
temoerature.
Safety Evaluation: The swapover logic from the lake to the ,
Standby Nuclear Service Water Pond was removed in the event of a LOCA. This also required a Technical Specification change since :
the cooling water for the plant will be supplied by the lake instead of the pond. This Technical Specification change will require daily monitoring of the lake temperature during the months July, August, and September. Currently there is a method to ,
measure RN Header temperature inlet to the NS Heat Exchanger.
This can only be done when the NS Heat Exchangers are being used.
- This change will relocate 1RNTE5000 and-1RNTE5010 upstream of Tee I branching to the NS Heat Exchangers. This modification will allow daily monitoring of the RN Header temperature during the months of July,' August, and September. This relocation will have a minimal affect on the flow in the RN Header due to the large diameter, 30 inches, of the pipe.
This modification will not increase the probability or coe sequences of an accident previously evaluated, or different than -
any already evaluated, in the FSAR. Nor will it increase the probability or consequences on an equipment malfunction, pre-viously evaluated, or different than any already evaluated, in the .
FSAR. The margin of safety defined in the bases of the Technical Specifications is unaffected. An unreviewed safety question does not exist. l
3 CE-1798
Description:
This Exempt Change will provide the necessary tubing tie-ins and isolation valves needed to easily obtain a high pressure final feedwater sample. The isolation valves will t,e @
under Chemistry control eliminating the need for assistance from l Operations. This change affects Figures 9.3.2-6 and 10.4.7-10 in the FSAR.
Safety Evaluation: During high pressure cleanup, a representative sample of the final feedwater cannot be easily obtained. Access to the sample point is not convenient and assistance from the Operations group is required.
The implementation of this change will not change the design or normal operation of this system. It simply provides a more convenient and accessible means of obtaining a sample during high pressure cleanup. The FSAR has been reviewed and this modifica-tion will not have any affect on the probability, con-sequences, or possibility of new or previously evaluated accidents. Also, this change will not affect the probability, consequences, or possibility of malfunctions of equipment important to safety evaluated in the FSAR. The margin of safety, as defined in the bases of the lech. Specs., will not be affected as related to offsite radiation doses and water volume requirements for the Condensate Storage System.
CE-1832
Description:
This change provides float traps for the Diesel CE-1833 Generator Starting Air Aftercooler shell to blowdown accumulated water. This change affects FSAR Figures 9.5.6-1 and 9.5.6-2.
Safety Evaluation: The Diesel Generator Starting Air Aftercoolers are collecting condensed water. Some of this water is being passed on into the drying tower. This condition is causing premature failure of the moisture separator and pre-filter traps and deterioration of the dryer absorbant. These conditions are causing the VG system to fail at an unacceptable rate. The intent of this modification is to provide Float Traps for the aftercooler shell to blowdown accumulated water. These traps will change the procedure from manual to automatic blowdown. Design has evaluated the seismic analysis and found it acceptable. Due to the auto-matic blowdown, the probability, possibility, or consequences of any accident or equipment malfunction is not increased by this modification. Implementation of this modification does not require any FSAR revisions as described in section 9.5.6.2.1. No Technical Specifications are affected by this modification.
Therefore, it is concluded that no unreviewed safety questions or concerns will arise as a result of this modification.
CE-1850
Description:
To allow one RN pump to supply the needs of both CE-1851 units in an accident scenario, the minimum RN flow to the KC heat exchangers (KC HX RN flow) was reduced to 5300 gpm. One result of the reduced KC HX RN flow is that the KC HX KC outlet temperature can exceed 120 F during a unit cooldown in the summer months.
120*F is the maximum KC HX KC outlet temperature operating limit to support the 40 year qualification of safety-related pumps
i 4
i cooled by KC. These motors were qualified for 40 years with normal cooling water inlet water temperature of 100*F 15'F with some excursions to 120*F.
per the one RN pump analysis, the KC HX KC outlet temperature can approach 130*F during a unit cooldown with a LOCA on the other unit in the summer months. To counter this temperature excursion and maintain a 40-year life span on the motors, a KC HX KC outlet temperature of 90*F 1 5'F for at least six months of the year (100'F i 5'r for the other six m;.nths) is needed, in order to accomplish the above, the KC HX KC outlet temperature setpoint will be lowered from 100'F to 90'F.
Safety Evaluation: The Component Cooling System (KC) is a closed loop system designed to supply cooling water to both safety equipment and r.an essential equipment. It operates during all phases of plant operation and >hutdown. The component cooling pumps and heat exchangers are arranged into two separate trains of equipment in each unit sub-system, with two pumps and one heat exchanger per train. The Nuclear Service Water System (RN) provides an assured source of cooling water to the component cooling heat exchangers.
The lowered temperature is in the conservatire direction and neither the lower temperature nor the change in velocities will adversely affect any equipment or system. The containment analysis for LOCA is also not adversely affected.
Since the system will function as before, the probability or consequences of an accident previously evaluated in the FSAR will not be increased. The temperature change cannot initiate an accident, so the possibility of an accident different than any already evaluated in the FSAR will not be created. Based on this change having no adverse affect on any equipment, there is no increase in the probability or consequences of a malfunction of equipment important to safety prevtously evaluated in the FSAR.
The new setpoint will not af fect any additional equipment nor is any new equipment added; therefore, the possibility of malfunc-tions of equipment important to safety different than any already evaluated in the FSAR will not be created. Since no safety para-meters or design limits have been adversely affe:ted, no margin of safety as defined in the bases to any Technical Specification is reduced. There are no unreviewed safety questions associated with these VN's.
CE-1937
Description:
This change revises the drain lines on the Lower CE-1935 Containment Ventilation Units 1A, IB, 10, and ID so that water CE-1936 will be routed out of the unit before going to the drain pan.
CE-1938 This change was initiated because there had been a problem with bearing failure on the fan motor due to water contamination. This change affects FSAR Figure 11.2.2-11.
2
e- i 5 l safety Evaluation: The modification of the drain lines on the !
Lower Containment Ventilation Units will not require any changes to the Technical Specifications for the Containment Ventilation System or the Liquid Waste Recycle System. This change will make the system more reliable because it will reduce the rate of bearing failure on the fan motor. This modification does not change the function of these systems.
No systems, structures or components important to safe operation of the plant will be affected by this modification. i CE-1859
Description:
This change revises flow diagrams to show the correct position'for conventional sampling lab valves during :
normal Conventional Sampling Lab operation. This change affects l FSAR Figures 9.3.2-6, 9.3.2-7, 9.3.2-8, and 9.3.2-10. ,
P Safety Evaluation: This change will show the proper normal operating position of Conventional Sampling valves on tystem flow diagrams. This will ensure that the Conventional Sampling flow diagrams are consistent with the Conventional sampling-cperating procedures.- The changes will also prevent the potential for >
system misalignments during tag outs.
These changes will not create an unreviewed safety question or require a change to the Technical Specifications.
CE-1880
Description:
IWLB49 is a normally open valve and should be locked open to prevent potential overpressurization of the associated Unit 2 drain header. This change revises FSAR Figure 11.2.2-13 to reflect this.
E Safety Evaluation: The safety functions of any system will not be affected by revising the affected document to show IWL849 as .
normally open and locked open. The draining function of the associated piping will not be affected. The existing procedu es used by the Operations Group already refor to 1WL849 as a normally !
open valve. If this valve were to be closed, a potential would exist for overpressurization of the associated Unit 2 WL system drain header. Locking the valve open will result in safer system operation, Revising the affected document as stated in this change will provide more accurate station documents.
'CE-1951
Description:
Currently, Low Pressure Service Water (RL) system Strainer B cannot be adequately isolated for maintenance due to excessive leakage past the inlet (IRL 32) and outlet (IRL 33) -
valves. To complicate the situation, 1RL 32 and 33 cannot be
- adequately isolated for repair due to excessive leakage past valves in the other low Pressure Service Water system headers.
This same problem could potentially develop at Strainer A due to :
identical design and components. This change was originated to install 3" drain lines upstream and downstream of the Low Pressure Service Water system strainers.
[
f b 6 Evaluation: The function or operability of the Low Pressure Service Water system will not be affected by installing these drain lines. The lines will be installed in accordance with the appropriate piping specification for this portion of the Low Pressure Service Water system. These drain lines will aid the draining process of the piping around the strainers so maintenance can be performed on the strainers. This variation noti:e does not increase the probability nor consequences of an accident pre-viously evaluated in the FSAR and it does not create the possi-bility of an accident not previously evaluated in the FSAR.
Installing these draio lines does not affect the function or operability of any safety-related equipment. Therefore, this change does not increase the probability nor consequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously evaluated in the FSAR.
For the same reasons, this variation notice will not reduce the margin of safety as defined in any Technical Specification basis.
CE-1957
Description:
Valve 1/2CA129 contributes to a reduction in the CE-1958 suction head for the auxiliary shutdown pumps. The valve internals will be permanently removed to reduce suction line pressure drop. This change will affect FSAR Figure 10.4.9-1.
Evaluation: This change provides for the permanent removal of the valve disc from 1/2CA129. The change will also note on the flow diagram that the disc has been removed. Removal of check valve internals is a typical maintenance activity covered by existing procedures. Valve 1/2CA129 is not required for system operation and is not part of the Pump and Valve In-Service Testing Program.
There will not be any functional tests required other than those performed for routine valve maintenance.
CE-1997 This modification replaces pressure switches 1/2KCPS5161 and CE-1998 1/2KCPS5211 with differential pressure switches. These switches alarm at 500 gpm difference between actual flow and demanded flow, This revision affects FSAR Figure 9.2.2-3.
Evaluation: This modification revises the Low flow alarm logic for the KC supply to the fuel pool heat exchanger. The low alarms are presently set at definite flow values. This design has created misance alarms as the needed Component Cooling water varies c.th spent fuel decay heat. The new alarm logic will use differential pressure switches that are set to cause an alarm at the flow setpoint minus 500 gpm. The recommended design setpoint for component cooling flow is 3000 gpm. This setroint is con-trolled by a manual loader and is specified in the I&C List.
Administrative Controls for the lower alarm setpoint will not allow flow to diminish to the point of starving the heat exchanger without alarming. This modification does not create the ability to inhibit proper fuel pool cooling. The consequences or pro-bability of an accident or malfunction of equipment previously
7 l
i evaluated in the FSAR will neither be increased nor created.- The margin of safety as defined in the basis to any of the Technical Specifications will not be reduced by this change. This modifica-tion does not create an unreviewed safety question.
CE-2005
Description:
This change deletes controls from valve 1/2CM841.
CE-2006 Steam Generator Blowdown Recycle Heat Exchanger flow will be controlled manually by valve 1/2CM56. Steam Generator Blowdown Recycle Heat Exchanger flow indication will be left operational.
This change affects FSAR Figure 10.4.7-5.
Evaluation: This change deletes the automatic controls on valve ICM841. This valve regulates the Condensate flow through the Steam Generator Blowdown Recycle heat exchangers. The controls for this valve have had continuous problems. Because of this the c valve has been failed open and flow through the Steam Generator Blowdown Recycle heat exchangers regulated by ICM56. This valve is down stream from ICM841 and is manually operated. This method of control has proved to be adequate. ICM841 has no function in plant transients. ICM56 can be throttled without damaging the valve. These are not safety related valves or a safety related system and this change to them will have no adverse impact on safety equipment. This is per the FSAR "any failure in the non-safety class portions of the Condensate and Feedwater systems does not prevent safe shutdown of the reactor."
No systems, structures, or components addressed in the FSAR will be affected in any significant manner by this change. Safety-related systems, structures, or components will not be degraded by this change. No changes to the FSAR or Tech. Specs, will be caused by this change. This modification will not cause any Unreviewed Safety Questions.
CE-2010
Description:
Nakeup Demineralized Water' flow diagrams do not reflect the actual valve positions of the following valves:
1/2YM341, 1/2YM342, 1/2YM343, 1YM410 and 1YM415. Also, 1YM415 is not properly located, This change updates the necessary drawings.
FSAR Figures 9.2.3-2 and 9.2.3-6 are affected.
Evaluation: Design documents of the Makeup Demineralized Water system do not reflect the as-built condition and the proper valve !
alignment. This variation notice was originated to revise the affected documents to reflect the following: the actual location -
of valve 1YM 415; valves 1/2YM 341,1/2YM 343,1/2YM 343,1YM 410 ;
and 1YM 415 as normally closed. Per the discussion below, this '
change does not involve safety-related equipment or an unreviewed safety question.
The function or operability of any system will not be affected by revising the affected documents. The Makeup Demineralized Water system evaluation in the FSAR is not af fected, Revising the documents will allow safer and more efficient operation of the Makeup Demineralized Water system. Therefore, this change does j not increase the probability nor consequences of an accident
-!u 8
previously evaluated in the FSAR and it does not. create the possibility of an accident not previously evaluated in the FSAR.
Revising the affected documents does not affect the function or '
[
operability of any safety-related equipment. Therefore, this variation notice does not increase the probability or con-sequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not.previously evaluated in the FSAR.
For the same reasons, this change will not reduce the margin of safety as defined in any Technical Specification basis. .
CE-2042
Description:
Instruments ONBPG6580, ONBPG6590, ONBPG6600 are deleted per this modification. The change affects FSAR Figures 9.3.5-3 and 9.3.5.
Evaluation: The deleted gages were suction pressure gages for the Boron Recycle Evaporator pumps. While the gages were in place, i chemistry personnel determined that they served no useful purpose '
for the operation of the pumps. The gages being deleted are not safety related. Removal of the gages will not affect the opera-tion of the pumps because it provides no useful operator informa-tion. The integrity of the system will be maintained because the root valves will remain in place and will be plugged. The proba-bility of an accident or malfunction of equipment evaluated in the FSAR will neither be increased or created by removing these gages. This change will not affect any system, structure or com-ponent addressed in the FSAR in a significant manner. No safety-related systems or components are being degraded by this modifica-tion. -
CE-2055
Description:
Valve 2SV64 was 1" Borg-Warner Y-Type globe valve.
This valve was replaced with a 1" Anchor-Darling Double Disc Gate Valve. This change affects F.
j Evaluation: Drain valve 2SV 064 is a 1" Borg-Warner Y-type globe-valve (Item No. 6J-207) and is located off the 6" main steam PORV header. This valve is not reliable and has a history of develop-ing leakage past the valve seat which results in main steam losses to the atmosphere. This change was. originated to replace the existing valve with a 1" Anchor-Darling double disc gate valve (Item No. 6J-335). Per the discussion below, this variation notice does not involve an unreviewed safety question.
Replacing valve 2SV 064 will not affect the function of the main steam system as evaluated in the FSAR. The new valve meets the functional and operational requirements of 2SV 064 and is a more reliable valve than the old one. This change does not increase the probability nor consequences of an accident previously eval-uated in the FSAR and it does not create the possibility of an i accident not previously evaluated in the FSAR.
- s. -
h 9
Replacing valve 2SV 064 will not affeet the function or opera-bility of any safety-related equipment. The new valve will perform the same function as the existing one and also wiil be less likely to develop leakage past the seat. Therefore, this variation notice does not increase the probability nor con-sequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously, evaluated in the FSAR.
For the same reasons, this change will not reduce the margin of safety as defined in any Technical Specification basis.
CE-2056
Description:
Valve 2SV67 was a 1" Borg-Warner Y-Type globe valve.
This valve was replaced with a 1" Anchor-Darling Double Disc Gate valve. This change af fects FSAR Figure 10.3.2-1.
Cvaluation: Drain valve 2SV 067 is a 1" Borg-Warner Y-type globe valve (Item No. 6J-207) and is located off the 6" main steam PORV header. This valve is not reliable and has a history of develop-ing leakage past the valve seat which results in main steam losses to the atmosphere. This variation notice was originated to replace the existing valve with a 1" Anchor-Darling double disc gate valve (Item No. 6J-335). Per the discussion below, this change does not involve an unreviewed safety question.
Replacing valve 2SV 067 will not affect the function of the SV system as evaluated in the FSAR. The new valve meets the func-tional and operational requirements of 2SV 067 and is a more reliable valve than the old one. This change does not increase the probability nor consequences of an accident previously eval-uated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
Replacing valve 2SV 067 will not affect the function or opera-bility of any safety-related equipment. The new valve will perform the same function as the existing one and will be less like'1y to develop leakage past the seat. Therefore, this change does not increase the probability nor consequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously evaluated in the FSAR.
For the same reasons, this change will not reduce the margin of safety as defined in any Technical Specification basis.
CE-2072
Description:
This change replaces the carbon steel piping down-stream of valve 1SM 131 with stainless steel piping. Also replace the remaining Schedule 40 carbon steel piping with Schedule 80 carbon steel piping.
Evaluatior.: The carbon steel drain lines from the main steam equalization header steam trap stations are developing leaks due to ercsion and corrosion. This change was originated to replace I
1 u
i 10 4
the carbon steel piping downstream of valve ISM 131 with stainless steel piping and to replace the remaining schedule 40 carbon steel piping with schedule 80 carbon steel piping. Per the discussion below, this variation notice does not involve an unreviewed safety question.
The function or operability of the Main Steam system will not be affected by the piping replacement. The piping and fittings installed will coroply with the appropriate piping specifications for this portion of the SM system. The stainless steel and schedule 80 carbon steel piping will be'less susceptible to erosion and corrosion and provide a more reliable piping system.
This change does not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR..
The piping replacement does not, affect the function or operability of any: safety-related equipment. The affccted Main Steam system !
-piping will not be degraded in any way. In fact, the new piping -!
will be more reliable. Therefore, this variation notice does not increase the probability nor consequences of safety-related equip-ment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related eqtipment malfunction not previously evaluated in the FSAR, No setpoints, operating parameters or safety limits are affected, so this variation notice will not reduce the margin of safety as defined in any Technical Specification basis.
CE-2088
Description:
This change revises FSAR Figure 9.5.7-2. The changes are: (1) delete "LC" (Locked Close) information for valve ILD119, and (2) delete the word " Locked" in Note 7.
Evaluation: This modification will remove the " locked closed" requirement from valve ILD119 and revise the Flow Diagram to reflect this information. The Operations group has thoroughly reviewed the locking requirement for this particular valve and determined it to be unnecessary. The reason code in the A7'(Valve Book) for locking ILD119 is number 50 which means " uncertain", j Based on this reason code and its obscure physical location, 4 Design Engineering has concurred with the locking requirement
{
deletion. >
l Valve IL0l' will continue to operate in the same manner as l before, i.e., open and close according to established operating [
procedures. For this-reason, the probability or consequences of '
an accident previously evaluated in the FSAR will not be in-creased. This valve is located beneath a barrier, requiring lifting equipment for access. Due to its location, the proba-bility of a malfunction of equipment important to safety will not be increased. Since a malfunction of equipment will not be .
increased, there should not be an increase in the consequences of equipment malfunction. The margin of safety as defined in the
. J
1 L
11 F
Bases to the Technical Specifications will not bt reduced since the valve will continue to operate as before using established F procedures. The answers to each of the 10CFR50.59 unreviewed safety questions are "no".
CE-2102
Description:
Additional drains were needed for the Nuclear ,
Service Vater piping between Component Cooling Heat Exchanger 1A J and valve IRN837 to facilitate heat exchanger cleaning. This change replaces the welded cap downstream of 1RNC10 with a blind flanged connection to provide an additional drain path for RN.
Evaluation: The purpose of this change is to provide an [
E additional drain for the Nuclear Service Water piping on the discharge side of Component Cooling HX 1A. The additional drain ,
is needed to support heat exchanger cleaning. Currently, there exists a 4" connection, off the Nuclear Service Water piping, with valve IRNC10, which is capped. This exempt change will replace i the welded cap downstream of IRNC10 with a blind flange. This will allow the 4" connection at 1RNC10 to be used as a drain path when this portion of RN is out of service. The flange will be !
installed downstream of 1RNC10 in non-safety piping. Design Engineering has reviewed this location and determined that it will '
not induce any harmful piping stresses. Although the Nuclear Service Water system is described in FSAR Section 9.2.1, the capped branch connection being modified is not addressed. This 4 modification has no impact on safety functionw during design basis events.
CE-2133
Description:
This modification corrects Instrument Air flow diagrams to agree with the as built condition of the plant.
Evaluation: This exempt change will update the affected Instrument Air flow diagrams to reflect field routed piping to corresponding instrumentatic9 root valves.
This change will provide for improved isolation to valves and
instruments using instrument air. Also, the plant can make more accurate isolation determinations when planning work. ;
This change does not adversely af fect any plant safety functions, so the consequences or probability of an accident or equipment malfunctions will not be increased. No new modes or operating characteristics are introduced by this change; therefore, the i possibility of new malfunctions or accident scenarios are not created. This change does not have any adverse impact on the margin of safety as defined in the bases to any Tech. Spec. ;
Therefore, based on this evaluation, no unreviewed safety question is involved with this change.
CE-2159
Description:
This change replaces the expansion joint upstream of the Nuclear Service Water Strainers with a spool piece.
a 12 Evaluation: The expansion joints upstream of the Nuclear Service i Water Strainers are susceptible to corrosion in the heat affected 3 zone for the welds joining the bellows and liner to the flanges. ;
Eventually. the expansion joints begin leaking and must '.>e replaced ;
or repaired. Weld repairs on the expansion joints are difficult i for the Station to perform due to the thinness of the bellows
(.050") and liner (.050"). . Replacement expansion joints will not alleviate repair difficulties. Also, replacements are expensive and not readily available. A spool piece is the preferred replace-ment since an expansion joint is not required.
per discussions with Design Engineering (DE), an expansion-joint is used in this application to compensate for piping misalignment and facilitate fit-up. Therefore, the expansion joint acts as a pipe to convey raw water and provide RN system integrity. A spool piece fabricated for this application will perform the same func-tion. The spool piece will be fabricated, tested, and installed ;
in accordance with Section XI of the ASME Code for Class III (Duke
- Class C) piping. This will ensure the spool piece is capable of maintaining RN system integrity. The spool piece will be easier to weld repair than the expansion joint because of the additional ,
thickness of the pipe wall. The pipe has a standard wall thick-ness of .375" compared to .050" for the expansion joint bellows.
DE has evaluated replacing the' expansion joint with a spool piece. '
DE-has determined that no undesirable stresses will be placed on piping, hangers, or other system components by this modification.
DE has also determined that no revisions to any hanger loads or settings are required and the seismic integrity of the Nuclear Service Water system will not be compromised by this change. ,
The function and operation of the Nuclear Service Water system will not be changed in any way by this modification. The RN system will still provide essential cooling during accident sequences, so the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Since RN is a mitigating system and not an accident initiator, this modification does not increase the pro-bability of any accidents as previously evaluated in the FSAR. As previously stated, the spool piece will meet ASME Code require-ments and Design has verified seismic integrity will be main-tained. Therefore, the possibility of an accident or malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created. For the same reasons, this modification will not increase the probability of a malfunt-tion of equipment important to safety previously evaluated in the FSAR. Since the function of the Nuclear Service Water system will not be impacted, this modification will not reduce the margin of safety as defined in the bases to any Technical Specification.
Based on the discussion presented, no unreviewed safety questions are judged to be involved or created by this modification.
IE l 13 CE-2197
Description:
This modification changes the Main Feedwater pump trips-on low suction pressure and flow from 5 seconds to 20 seconds.
Evaluation: This change extends the Main Feedwater pump trip time ,
delay for low. suction flow and pressure. The time delay is being l extended from 5 seconds to 20 seconds. The pump manufacturer and Design Engineering have reviewed this change and agreed to its implementation. Reference attachments. The worse case accident that could result from these time delays being extended would be a.
loss of the pumps. However, these pumps are not safety related and per the FSAR 10.4.7.3. "Any failure of the non-safety class portions of the condensate and feedwater systems does not prevent safe shutdown of the reactor." The consequqnces or probability of an accident or malfunction of equipment evaluated in the FSAR will-neitner be increased nor created. The margins of safety as e defined in the basis to any of the Tech. Specs, will not be reduced. -This modification does not create an unreviewed safety
_ question.
CE-2202
Description:
This modification removes the wash fountains from !
the hot side of the men's and women's change room. This change affects FSAR Figure 6.3.2-3.
Evaluation: This modification removes the two 36" dia, wash fountains in the men's hot side of the change room and the 54" dia, wash fountain in the women's hot side of the change room.
The removal of-this equipment will require Liquid Waste Recycle drain pipe to be capped.
This change is non-QA Condition. This change does not affect or
. negatively impact any system, structure, or component necessary to .
the safe operation or safe shutdown of the plant. Therefore, the l probability, possibility or consequences of any accident or safety related equipment malfunction is not increased by this modifica-tion.
CE-2208
Description:
The Diesel Generator annunciator has alarms for "High AP Fuel TP Strainer 1" and "High AP Fuel TP Strainer 2". l The pressure switches which provide the alarm have been deleted.
This change affects section 7.6.16.1 of the FSAR.
Evaluation: The implementation of this change does not affect the system in any way. It removes the annunciator windows for "High AP Strainer 1" and High AP Fuel TP Strainer 2". The pressure switches which provide these alarms have already been removed.
The system still functions as designed. <
This modification will not increase the probability or con-sequences of an accident previously evaluated or different from any.already evaluated in the FSAR. It will not increase the probability.or consequences of an equipment malfunction previously evaluated in the FSAR. The margin of safety as defined in the bases of the Technical Specifications is unaffected. There is no unreviewed safety question.
E 14 CE-2253 Descriptioni The test connection for Steam Generator Power Operated Relief Valve is too large. This change replaced the existing 1" connection with a 1/2" connection.
L Evaluation: ~In order to perform post modification testing for NSM t
CN-50395, Rev. O, an instrument test connection is needed at the valve bonnet drain piping for Steam Generator (S/G) Power Operated Relief Valve (PORV) ISV013. This connection will normally be plugged and only used for testing purposes. This change was originated to provide the test connection at the one inch diameter bonnet drain piping for valve ISV013. Per the discussion below, an unreviewed safety question does not exist.
, The function or operability of.S/G PORV ISV013 will not be affected by providing the instrument test connection at the bonnet drain piping. The Main Steam system will continue to function as evaluated in the FSAR. The affected drain piping will continue to drain the valve bonnet area as originally designed. The new instrument test connection will meet the pressure and temperature requirements of this portion of the SV system. Per Design Engineering, the added weight of the test connection does not create any seismic concerns. Therefore, this exempt change does I not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
As stated above, the function of S/G PORV ISV013 and its ;
associated bonnet drain piping will not be degraded.. Providing the test connection specified in this exempt change will not affect the function of any safety-related equipment. Therefore, this exempt change does not increase the probability nor con-sequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously ayaluated in the FSAR.
CE-2254 Descripti6n: This change provides an instrument test connection CE-2255 .at the 1" Steam Generator Pressure Operated Relief Valve bonnet drain piping for valves 1/2SV001, 1/2SV007, and 1/2SV019.
Evaluation: In order to perform testing of Steam Generator (S/G)
Power Operated Relief Valves (PORV) 1/2SV001,1/2SV007 and 1/2SV019, an instrument test connection is needed at the valve bonnet drain piping. This connection will be used to measure the PORV bonnet area pressure at test conditions. This connection ,
will normally be plugged and only used for testing purposes. The testing will determine the operability of the above PORV's. The valve manufacturer, which is Control Components, Inc. (CCI),
released information concerning random failures of these valves so the decision was made to test the valves for operability. This exempt change was originated to provide the test connection at the one inch diameter bonnet drain piping for the above valves. Per the discussion below, an unreviewed safety question does not exist.
.i
15
.The function or operability of $/G PORV's 1/2SV001, 1/2SV007 and 1/2SV019 will not be affected by providing the instrument test connection at the bonnet drain piping. The Main Steam system will continue to function as evaluated in the FSAR. The affected drain piping will continue to drain the valve bonnet area as originally designed. The new instrument test connection will meet the pressere and temperature requirements of this portion of the SV system. per Design Engineering, the added weight of the test connection does not create any seismic concerns. Therefore, this exempt change does not increase the probability nor consequences of an. accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
As stated above, the function of the above S/G PORV's and their associated bonnet drain piping will not be degraded. Providing the test connection specified in this exempt change will not affect the function of any safety-related equipment. Therefore, this exempt change does not increase the probability nor con-sequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously evaluated in the FSAR.
CE-2347
Description:
This modification is for the installation of flow orifices in_the main feedwater system. Due to restricted flow and excessive cavitation, flow orifice 2CFFE 6330 which was replaced by a previous modification must be changed back to the original 1 orifice for steam generator loop C. This change affects FSAR Figure 10.4.7-11.
The original flow orifice was a multihole orifice and was in unit operation until implementation of a previous modification.
However, the original orifice did impede system operation by limiting unit operation at less than 100 percent due to similar flow restrictions, but without the cavitation.
The implementation of this modification will not require any changes to the Technical Specifications or FSAR for the Feedwater (CF) System. This modification will not create any unreviewed safety questions since the CF Systems function is not being changed, and since the unit will be in a mini-outage and power <
reduction / escalation is not a problem. Design has all original l calculations which qualify the multihole orifice for use in the CF j System, i Since the unit will be in a mini-outage the results of changing this orifice do not reduce any margin of safety to any safety-related system, component or equipment.
This modification will not require any changes to the Technical Specification or FSAR for Catawba Nuclear Station, and the ability of the Main Steam system to function as before as a result of this modification is not af fected.
l
F 1 16
)
Neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased, since the flow
. orifice was designed and functioned in the CF System prior to ;
removal. Since the orifice will function as before, no possi-bility of an accident is being created which is different than already evaluated in the FSAR. Also, since the system operating parameters are not being changed, the probability nor the con- !
sequences of a malfunction of equipment important to safety d previously evaluated in the FSAR will not be increased. Based on t the above, the system operating parameters not changing, possible malfunctions of equipment important to safety different than f previously evaluated in the FSAR will not be created.
No safety parameters or design limits are being affected and no margin of safety as defined in the bases to any Technical Specification is reduced.
CE-2263
Description:
This change revises the applicable drawings to allow '
for the replacement of vendor supplied gate valves on the Nuclear Service Water Supply lines to the Lower Containment Ventilation Unit 20.
This change will revise the applicable drawings to allow for the replacement of leaking 3/4" Vendor Supplied Gate Valves on the Nuclear Service Water Supply lines to Lower Containment Vent 11a-tion Unit 20. These valves are normally closed and when opened drain this portion of Nuclear Service Water into the Liquid Waste Recycle, and by these valves constantly leaking the additional flow to WL is increasing the amount of WL which has to be pro-cessed unnecessarily.
[
This modification of the drain valves on the Nuclear Service Water
- supply lines on Lower Containment Ventilation Unit 2D will not require any changes to the Technical Specifications or FSAR for the Nuclear Service Water System (RN), or the Containment Ventila-tion System (VV). This modification will not create any unreviewed ,
safety questions since the VV and RN systems function are not being changed. Because the systems will be isolated, and there is no change being made to the VV and RN systems function, the chances of an accident will not increase. Neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased, since the VV and RN systems will continue to function as evaluated in the FSAR. No new accident will be created since the affected portion of the VV and RN. systems will
- be out of service during this modification. Also, the probability or consequences of a malfunction of safety-related equipment will not be increased because the function of the VV and RN systems are unaffected by this modification. For the same reason, no new possibility of malfunction of safety-related equipment is created and replacement of the drain valves will not affect any margin of safety to any Technical Specification.
L i 17 i
CE-2262
Description:
This change revises the applicable drawings to 4110w for the replacement of vendor supplied gate valves 2RNA35 through ,
E 2RNA42 on LCVU20. This change affects FSAR Figure 9.2.1-11.
Evaluation: This modification of the drain valves on the Nuclear I Service Water supply lines on Lower Containment Ventilation Unit 20 will not require any changes to the Technical Specifications or FSAR for the Nuclear Service Water System (RN), or the Containment
- Ventilation System (VV). This modification will not create any unreviewed safety questions since the VV and RN systems function t are not being changed. Because the systems will be isolated, and there is no change being made to the VV and RN systems function, the chances of an accident will not increase. Neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased, since the VV and RN .
systems will continue to function as evaluated in the FSAR. No [
new accident will be created since the affected portion of the VV '
and RN systems will be out of service during this modification.
Also, the probability or consequences of a malfunction of safety-related equipn.ent will not be increated because the function of the VV and RN systems are unaffected by this modification. For i the same reason, no new possibility of malfunction of safety- f' related equipment is created and replacement of the drain valves will not affect any margin of safety to any Technical !
Specification.- l i
CE-2261
Description:
The change revises the applicable drawings to allow j for the replacement of vender supplied gate valves. 2RNA21 :
through 2RNA28 on Lower Containment Ventilation Unit 2B. These [
valves leak by into the Liquid Waste Recycle System. This change i affects FSAR Figure 9.2.1-1. l Evaluation: This modification of the drain valves on the Nuclear Service Water supply lines on Lower Containment Ventilation Unit 2B will not require any changes to the Technical Specifications or FSAR for the Nuclear Service Water System (RN), or the Containment Ventilation System (VV). This modification will not create any unreviewed safety questions since the VV and RN systems function are not being changed. Because the systems will be isolated, and !
there is no change being made to the VV and RN systems function, the chances of an accident will not increase, Neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased, since the VV and RN 1 l systems will continue to function as evaluated in the FSAR. No l new accident will be created since the affected portion of the VV i and RN systems will be out of service during this modification, Also, the probability or consequences of a malfunction of safety-l related equipment will not be increased because the function of i l the VV and RN systems are unaffected by this modification. For l the same reason, no new possibility of malfunction of safety- ;
related equipment is created and replacement of the drain valves will not affect any margin of safety to any Technical Specification.
l 1
y 18 l
U
( .
j
Description:
This change revises the applicable drawing to allow !
for replacement of vendor supplied gate valves 2RNA07 through !
2RNA14 in Lower Containment Ventilation Unit 2A. These valves leak by into the Liquid Waste Recycle System. The change affects FSAR Figure 9.2.1-1. l Evaluation: This modification of the drain valves on the Nuclear Service Water supply lines on Lower Containment Ventilation Unit 2A will not require any changes to the Technical Specifications or ;
FSAR for the Nuclear Service Water System (RN), or the Containment '
Ventilation System (VV). This modification will not create any unreviewed safety questions since the VV and RN systems function are not being changed. Because the systems will be isolated, and !
6 there is no change being made to the VV and RN systems function, the chances of an accident will not increase. Neither the proba-
- bility nor the' consequences of an accident previously evaluated in the FSAR will be increased, since the VV and RN systems will ,
- entinue to function as evaluated in the FSAR. No new accident !
will be created since the affected portion of the VV and RN systems will be out of service during this modification. Also, !
the probability or consequences of a malfunction of safety-related equipment will not be increased because the function of the VV and RN systems are unaffected by this modification. For the same reason, no new possibility of malfunction of safety-related .
equipment is created and replacement of the drain valves will not i affect any margin of safety to any Technical Specification.
CE-2274
Description:
Valve 2CA242 had excessive metal removed from the seat during the lapping process. Since the valve cannot be repaired, a new valve must be installed and one with the same model number is not available. This change updates the necessary drawings.
Evaluation: The replacer;.ent of 2CA242 with an Anchor / Darling double disc gate valve will not change the normal operation of the CA system. Valve 2CA242 is a 1 inch drain valve which is not required to be operated during normal or accident situations. The replacement valve is heavier than the existing valve, however, Design Engineering has reviewed this relative to their piping stress analysis and said the additional weight was acceptable.
There are also differences between the two valve design para-meters. The flow characteristics are insignificant since this is only used for drain purposes. The temperature and pressure rating of the new valve are slightly less than the existing valve, but-still exceed the design parameters for this application. The size of the new valve is similar to the Borg-Warner valve. The guide- .
lines on CN-1680-47 will still be met with the installation of the !
Anchor / Darling valve. No changes to the FSAR or Technical Specifi-cations will be required because of this modification.
CE-2308
Description:
Since eddy current testing did not provide a complete examination of thimble C7, this thimble needed to be isolated for unit startup as a conservative measure. This change provides a QA qualified fitting to cap off the thimble and updates the affected .
drawings as necessary. I L
19 Evaluation: This exempt change revises affected documents to show the following: Unit 2 Incore Detector System Flux Thimble C-7 sealed off at the seal table with a QA-1 Swagelok cap fitting, the removal of tube and fittings between Thimble C-7 and its isolation valve, and isolation valve for Thimble C-7 closed off.
The need to seal off thimble C-7 on the Unit 2 Incore Detector System was discovered during Eddy Current Testing for thimble wall thinning conducted at V2EOC2. Eddy Current Tests were conducted to assess the magnitude of vibration induced thimble wall thinning which was identified as an industry wide concern in Westinghouse reactors by NRC Bulletin 88-09. The Eddy Current Test conducted on Unit 2 Thimble C-7 did not provide a complete enough examina-tion of thimble C-7 to ensure that no wear is present. As a conservative measure, the decision was made to cap Unit 2 Thimble C-7 until further Eddy Current Test can be conducted in the future.
, The need to isolate Unit 2 Thimble C-7 was identified as a pre-requisite for unit startup. The isolation valves originally supplied for each of the 58 incore detector flux thimbles (on both Unit I and 2) are not Nuclear Safety Related Components. In the event of a breach of Unit 2 Thimble C-7 integrity (in an area of through wall wear) during unit operation, the isolation valve at the seal table would become part of the Reactor Coolant System pressure Boundary. Not being a Nuclear Safety Related Component, this valve would not be qualified to fulfill this function.
Therefore, the decision was made based on discussions with a cognizant Westinghouse representative to remove the tube and fittings between Unit 2 Thimble C-7 and its isolation valve, close off its isolation valve, and cap Unit 2 thimble C-7 at the seal table with a QA-1 qualified cap fitting. This seal would serve as an acceptable component of the Reactor Coolant Pressure boundary in the event of failure of the thimble.
This change will not increase either the probability or the consequences of any accident previously evaluated in the FSAR.
The unlikely event of Thimble C-7 leaking due to excessive through wall wear and the cap leaking at the seal table has been evaluated from the standpoint of a small break 1.0 FT LOCA, Section 15.6.5 of the FSAR. The isolation of leaking thimbles is addressed specifically in Section 7.7.1.9 of the FSAR.
The possibility of an accident or equipment malfunction not previously analyzed in the FSAR will not be increased by this change. The margin of safety as defined in the applicable Tech.
Specs will not be degraded by this change. The bases of Tech.
Specs. 3/4.2.1, 3/4.2.2, 3/4.2.3, and 3/4.2.4 are all critically affected by availability of the incore detector system. Thii
. change does not reduce the instrumented core locations to a level of endangering system operability.
For the above reasons, no unreviewed safety questions exist.
p 20 i
CE-2348
Description:
Flow orifice 2CFFE6340 needed to be replaced with I the original orifice because of excessive cavitation.
Evaluation: This modification is for the installation of flow orifices in the CF Feedwater System. Due to restricted flow and i excessive cavitation, flow orifice 2CFFE6340 which was replaced by i NSM CN-20545 requires changing back to the original orifice for
- i. steam generator loop 20. ,
The original flow orifice was a multihole orifice and was in unit operation until implementation of a station modification. ,
However, the original orifice did impede system operation by 4
limiting unit operation at less than 100 percent due to similar L flow restrictions, but without the cavitation.
The implementation of this modification will not require any j changes to the Technical Specifications or FSAR for the Feedwater '
(CF) System. This modification will not create any unreviewed ,
safety questions since the CF Systems function is not being I
changed, and since the unit will be in a mini-outage and power reduction / escalation is not a problem. Design has all original calculations which qualify the multihole orifice for use in the CF System.
This modification is QA Condition 4 and since the unit will be in a mini-outage the results of changing this orifice do not reduce any margin of safety to any safety-related system, component or equipment.
This modification will not require any changes to the Technical Specification or FSAR for Catawba Nuclear Station, and the ability of the CF system to function as before as a result of this modification is not affected.
Neither the probability nor the consequences of an accident previously evaluated in the FSAR will be increased, since the flow .
orifice was designed and functioned in the CF System prior to removal. Since the orifice will function as before,- no possi- i bility of an accident is being created which is different than !
, already evaluated in the FSAR. Also, since the system operating parameters are not being changed, the probability nor the con-sequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Based on ,
the above, the system operating parameters not changing, possible .
malfunctions of equipment important to safety different than previously evaluated in the FSAR will not be created.
No safety parameters or design limits are being affected and no margin of safety as defined in the bases to any Technical Specification is reduced. '
It is concluded that an unreviewed safety question does not exist.
?
m _ _ _ - _.
21 p
CE-2397
Description:
This change removes Auxiliary Steam valves SA-16 and SA18 and is replacing them with piping and pipe unions. The pipe clan break locations will also be changed. In addition, valve SA144's valve bonnet will be oriented so that the drain time connection comes off the bottom and the drain line will be rerouted. This change affects FSAR Figure 10.3.2-1.
< Evaluation: SA-16 is a valve in the Auxiliary Feedwater Pump Turbine Governor Valve Stem Leakoff Line, which drains to the L Steam Turbine Driven Auxiliary Feedwater Pump 1 Sump. SA-17 and SA-18 are valves in the Turbine Exhaust drain lines of the AFPT.
The function of these lines is to remove condensate and they are closed only during maintenance operations. Removal of SA-16, SA-17 and SA18 will not adversely affect Main Steam Supply to Auxiliary Equipment (SA) System operations and will have no effect on stress analysis (Ref 4 and 5). The reorientation of the drain line connection and the drain line reroute will have no adverse effect on the operation of either SA-144 or the SA system (Ref.
7). The use of unions has been evaluated and found to be acceptable (Ref. 4 and 8).
Removal of these valves will require reclassification of the pipe .
in the area of the AFPT from Class B to Class C (for details see [
Attachment 2). The only portion of the SA System required to be !
Class B is that portion that serves a containment isolation i function. Therefore, since the pipe in the area of the AFPT that ;
is being downgraded is not associated with containment isolation, l the downgrading of the pipe to Class C is acceptable. j Since the affected system will function _as before and no FSAR accident initiators are altered, the probability of an accident or malfunction of equipment important to safety previously evaluated in the FSAR is not increased. Since the drain lines will still perform their function and the operation of all systems including SA is not adversely affected, the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
No new failure modes are added by this work. Therefore, the possibility of an accident or malfunction of equipment important !
to safety, different than already evaluated in the FSAR, is not created. l No plant parameter, setpoint or safety limit is altered by this l work. Therefore, the margin of safety as defined in the bases to !
any Technical Specification is not reduced. There are no USQs ;
associated with the work as defined in the Description of Work. 1 l
l CE-2409
Description:
Valve IRY224 is stuck-in the open position and can not be used for an isolation valve for the Hose Rack in the Waste Solidification Building. This valve will be shown on the flow 'i diagram as non-functional. The QA3 boundary will be relocated i
i
22 from this valve to the Hose Rack Valve (IRF298). 1RF298 will be used as an isolation valve for the Hose Rack. Roadway box cover i for 1RY224 will be tack welded shut to prevent future operation of the valve. This change affects FSAR Figures 9.5.1-1 and 9.5.1-6. ,
Evaluation: An underground valve in the fire protection system (IRY224) located behind the Waste Solidification Building is stuck in the open position. This valve is required by Tech. Specs.
(3/4.7.10) to be cycled at least once per 12 months, and its correct position verified at least once per 31 days. Repeated attempts to cycle this valve have failed (Ref. WR#50565 OPS). -
This exempt change was originated to revise the flow diagrams and piping layout drawings to show this valve as non-functional, and ;
therefore, non-testable. Per the following discussion, an unreviewed safety question does not exist.
By calling this valve non-functional, the following flow diagram changes result:
- The QA condition 3/Non QA line break is relocated from 1RY224 to the valve at the Hose Rack (1RF298) in the Waste -
Solidification Building. -
- Valve IRF298 becomes the isolation valve for this Hose Rack.
- Piping and components between IRY224 and 1RF298 are '
upgraded to QA candition 3.
Design Engineering has performed a justification for upgrading the piping and components to QA condition 3. Tests and inspections conducted and documented under WR#50565 OPS have verified that 1RY224 allows adequate flow to meet design basis commitment for the fire protection water distribution system; and that QA condition 3 criteria are met for the upgraded portion of the piping. 1RY224 was used to isolate only one Hose Rack in the t Waste Solidification Bldg. The ability to isolate the Hose Rack is still intact by use of 1RF298. Valve 1RY224 will be protected from further operating by welding shut the associated roadway box ,
cover. The function or operability of the RY/RF system will not be affected by the changes discussed above. Therefore, this exempt change does not increase the probability nor consequences of safety-related equipment malfunction previously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction not previously evaluated in the FSAR. For the same reasons this exempt change will not reduce the margin of safety as defined in any Technical Specification bases. _
Additionally, this exempt change does not increase the probability nor consequences of an accident previously evaluated in the FSAR and it does not create the possibility of an accident not previously evaluated in the FSAR.
1 l
e _
i b CE-2415
Description:
Valve 2SV65 was a 1" Borg-Warner Y-Type globe valve.
- 2SV65 was replaced with a 1" Anchor-Darling Double Disc Gate F Valve. This change affects FSAR figure 10.3.2-1.
! Evaluation: Drain valve 2SV065 is a 1" Borg-War.,er Y-type globe l valve. This valve is located off the 6" power-operated relief valve (PORV) header in the interior doghouse. The Borg-Warner valve is not reliable and has a history of developing leakage past the valve seat which results in main steam losses to the l' atmosphere. This exempt change was originated to replace the existing valve with a 1" Anchor-Darling double disc gate valve.
Replacing valve 2SV065 will not affect the function of the Main Steam Vent to Atmosphere (SV) system as evaluated in the FSAR.
The draining function of the valve will not be affected. The design of the new valve meets the temperature and pressure condi-tions experienced in this portion of the SV system. The new valve will perform the same function as the existing one and also will be less likely to develop leakage past the seat. Per Design Engineering, the additional weight of the new valve does not create any seismic concerns or require any additional support.
CE-2477
Description:
Existing valve 1YC342 cannot be repaired and no replacement valve is available. This valve will be changed to a 1/2" Ball valve. This change affects FSAR Figure 9.4.1-6.
Evaluation: This Exempt Change will revise drawings to allow for the installation of a new low point drain valve (1YC342) on the YC System. The existing valve _cannot be repaired and will be re-placed by a 1/2" Grinnell Ball Valve. This modification is QA Condition 1.
This modification is required to be completed before 9-27-89 in order to adhere to Technical Specification 3/4.7.6 which requires that both trains of YC be operable during Modes 1, 2, 3 and 4.
The existing valve 1YC342 is leaking and does not contain the YC system boundary thus making entry into the 7 day action statement on YC.
This modification will be performed while the affected portion of !
the YC System is out of service and will not impact any other system important to the safe operation of the station. l The implementation of this change will not require any changes to-the Technical Specifications or FSAR's for the YC system, and will not create any unreviewed safety questions since the YC systems-function is not being changed, and the chances of an accident.not previously identified will not be increased since the YC system will continue to function as evaluated in the FSAR. !
i No new accident will be created since the affected portion of the j YC system will be out of service. Also, the probability or consequences of malfunction of safety-related equipment will not
]
g _
.z i
' {u h gy '4' "
k .< 24;
-1 t
+
1 h - -- -
, beiincreased because the functionLof the YC. system is. unaffected '
, , - by1this-modification. For'the same reason,'no new possibility of f.l' ' ; malfunction of safety-related equipment is created and installa-
-tion of'the new: valve lwill not affeet any margin:of safety to.any-Q g - Technical-Specification, o "f ;- .
It~is concluded that-no unreviewed safety question or. concern will. !
'1 ' arise as a: result of the implementation of:this Exempt Change. :
j n ';
1
(
)? f f
- g 1 3
e 1
- ll -- f
!. {
t
> i I! .' -
\!$
,(
i
'fi- e i i $. $
1}
L s
s -l 3
5
. [
CATAWBA NUCLEAR STATION
SUMMARY
OF PROCEDURE CHANGES - TESTS, AND EXPERIMENTS COMPLETED UNDER 100FR50.59
- ./R MU0/06 Mirror and-Blanket Insulation Removal, Modification and Reinstallation: .This is an original procedure issue to document and describe insulation mainte-nance activities. Duke Power Design Engineering performed a study of all pertinent potential safety concerns, including such issues as a calculation to-determine the quantity of insulation debris which could be generated by a worst case high energy line break vs, containment sump intake screen face area, etc. No unreviewed safety questibns (henceforth USQ) were deemed to exist.
MP/0/A/7650/107 Retubing KC Heat Exchanger: This is a procedure reissue to add several sign-offs for leak testing and tube hole inspection. The changes were only editorial in nature; the procedure content remained the same, No USQ was deemed to exist.
MP/0/A/7650/86 Functional Testing of Mechanical Snubbers: This is a procedure reissue which adds information and data to facilitate testing of Anchor / Darling mechanical snubbers, The testing performed ensures operability and performance to design specs and manufacturers' requirements. No USQ was deemed to exist.
MP/0/A/7650/66 Auxiliary Feedwater Pumps Carbon Dioxide Cylinder Weight Verification:- This is a procedure reissue which adds' steps to ensure proper isolation prior to performing work, steps to ensure the proper condi-tion of braided hoses, steps to ensure the proper condition of manifold threads, and steps to ensure the proper condition of other parts, components, and valves. These steps will enhance safety and ensure performance to design specs. No USQ was deemed to exist. ,
PT/1/A/4400/01 ECCS Flow Balance; Change #12: This change adds a step to place an electrical jumper to bypass the interlock that requires INI36B to be closed in . order to open 1NI1478. Opening INI1478 with INI368 also open could result in reducing reactor coolant system inventory, with the water. going to the refueling water storage tank. However, to prevent this occurrence, this change also requires that INI36B be closed with power removed. Therefore, no USQ was deemed to exist.
PT/0/A/4400/08 RN (Nuclear Service Water) Flow Balance for Degraded ,
Mode, Change #28: This procedure change allows RN d
I !
v
'I pumphouse flows to be balanced without completing an entire RN system flow balance. This is conservative because.the RN pump discharge pressure will be lowered to a pressure at or below that of the previous RN system flow balance. This will make flows achieved in the pumphouse comparable to those ;
cf a full RN system flow balance. The minimum flows utilized in the procedure to perform the flow balance have been reviewed by Design Engineering and -
the RN pump manufacturer. No USQ was deemed to exist.
PT/1/A/4400/01 ECCS Flow Balance, Change #8: This change incorpo- #
rates recent flow balance criteria analyzed and provided by Duke Power Design Engineering. The flow balance is performed with no fuel in the core (no mode), so no affects on core parameters will ensue as a result of the performance of this test. This 1 change also corrects a typographical error and adds independent verification steps to ensure proper :
installation and removal of test equipment.
PT/2/A/4200/09A Auxiliary Safeguards Test Cabinet Periodic Test, Change #56: This change includes provisions to "
ensure that the other side of the penetration is isolated when certain nuclear sampling containment isolation valves are stroked. This will ensure that-reactor coolant system pressure is isolated from the sample header during the test, thereby enhancing huclear safety. This change also corrects a typo-graphical error. No USQ was deemed to exist.
PT/1/A/4350/15B Diesel Generator IB Periodic Test,-Change #15: This '
change allows the installation of temporary 120VAC power to the Lo-Lo lube oil transmitter located in cabinet 1ELCP0329 for D/G 18. Normally, the D/G provides power while in operation. Part of this procedure is performed with the D/G shutdown and Lo-Lo lube oil contacts and circuitry being tested.
During this test, the D/G is inoperable. Therefore, "
l
' the use of a non-safety related system power supply 1s of no consequence. This change provides for i thorough test while limiting starts and run 'ime for D/G 18. Because of-the above, the probabil t/ of an
- equipment malfunction is reduced. No USQ was deemed to exist.
PT/1/A/4350/15A PT/1/A/4350/15B Diesel Generator 1A (IB) Periodic Test, Changes #14 and #16: NSM CN-11104 replaces the D/G.1B Lo-Lo Lube Oil and Overspeed emergency trips with elec-tronic trips. After experiencing problems with the original mod, Solonoid 99 was added in the pneumatic circuitry to ensure that non-emergency trip signals 2
i.
will not trip D/G IB under emergency start condi-tions. These changes add acceptance criteria and steps to test Solonoid 99. The method of testing j Lo-Lo Lube oil trip circuitry has also been changed i to ensure that all possible failure possibilities !
have been tested. !
The modifications to Diesel Generator 18 will be performed under NSM CN-11104 and associated work requests. The diesel generator will be tagged out and will remain inoperable until all modifications - +
and testing has been completed. _The opposite train will remain operable and capable of performing any-required safety functions. Only one diesel genera- !
tor is required in modes 5 and 6. Therefore, the-probability or consequences of any new or previously ;
evaluated accident will not be increased. The probability or consequences of equipment malfunction will not be increased. The margin of safety will not be reduced. No USQ was deemed to exist.
PT/1/A/4200/13H NI (Safety Injection) and NV (Chemical and Volume '
Control) Check Valve Test, Change #9: This change will allow the test to be performed without making either train of ND (Residual Heat Removal) inopera-ble, since fuel will be in the core during perfor-mance of this procedure. Per this change, two trains of and will always be operable, constituting an enhancement of nuclear safety. This change will allow an ND pump to supply two cold legs with 3000 GPM and supply NI pump hot leg injection and NV pump ;
cold leg injection simultaneously.
Preoperational testing shows that in this alignment,-
ND pump cold leg injection flow will be about 3000 GPM, total NV pump flow about 700 GPM, and total NI pump flow of about 1300 GPM for a total ND pump flow "
of about 5000 GpM, which is below the pump run out value of 5500 GPM. No USQ was deemed to exist.
PT/1/A/4200/10B Residual Heat Removal Pump 1B Performance Test, Change #46: This procedure change allows Residual-Heat Removal (ND) Pump 18 to be tested taking suction of f of NC Loop C, instead of the Refueling -
Water Storage Tank. The opposite ND train (A-Train) will be in operation during the test, as required by Technical Specification 4.9.8, thus insuring ade-quate core cooling. Pump Miniflow will still be' employed, and leakage past IND60 will be minimal, so as not to overcool the core. Additionally, the Control Room Operators may interrupt this test at any time should plant parameters dictate so. The valve alignment to be used will be based on the Normal Cooldown alignment contained in OP/1/A/6200/04, 3
Enclosure 4.1, with the exception that IND53 must be available for throttling should pump flow be too i
high, Following the pump test, if IN053 has-been throttled, it will be. returned to full open to insure pump operability. No USQ was deemed to exist.
PT/1/A/4150/04
_ PT/2/A/4150/04 Reactivity Anomaly Calculation, Procedure Reissue:
Reissue of this procedure slightly affects its function with regard to validation of predicted total core reactivity by adding guidance for ad-
-justment of core reactivity for core Tavg/T-Ref error. The basis for this procedure-is stated in -
Tech Spec 4.1.1.1.2 which requires such a measure-ment to ensure adequate shutdown margin exists in modes 3 and 4 Additionally, performance of this procedure validates the design assumptions with respect to assured shutdown margin in mode 5 (Tech Spec 3.1.1.2). It also irdirectly ensures unit operation in compliance with required moderator-temperature coefficient (Tech Spec 3.1.1.3) which has dependence upon Reactor Coolant System boron concentration. No change to any Tech Spec will be required as a result of this change.
Reissue of this procedure has not effect on any procedure mentioned in Section 14.0 of the FSAR. It does affect a test (Reactivity Anomaly Calculation) which is not mentioned in the FSAR. However, due to the routine nature of this test (being for the purpose of periodic surveillance) it does not require inclusion in the FSAR. This is the inter-pretation taken from Part 7.1.0 of " Guidance for 10 CFR 50- 59 Evaluations" issued by Duke Power Company.
This procedure does not affeet structures,-systems.-
or components that are addressed in the FSAR in a significant manner. The only affected system is the Nuclear Sampling (NM) System which is employed in a routine manner, under the control of Chemistry Group procedures, to draw a reactor coolant sample.
Analysis of this sample to determine reactor coolant boron concentration is then performed.
The probability of an accident previously evaluated in the FSAR will not be increased by the reissue of this procedure. The potential impact of acquisition of NC System boron samples required by this proce-dure is bounded by the analysis of Section 15.6.2, Break in Instrument Line or other Lines from Reactor Coolant Pressure Boundary that Penetrate Contain-ment, which considers use of the Hot Leg Sample Lines.
4
p B
V , ,
p
~
Adequate shutdown margin.is verified by the-periodic performance of this procedure. This surveillance- '
assures that the core reactivity throughout core life has been accurately modelled. The core design-t
.is thereby verified'to be in compliance with the required shutdown margin of 1.3% AK/K in modes 3 and 4 and 1.0% AK/K in mode 5.
Failure of this procedure to demonstrate that the calculated core reactivity is within 500 pcm of the I. predicted value would inithte ? review to determine t whether'or not an investigation into the validity of ,
the assumptions upon which the core design was based is required. This would preclude the possible violation of the shutdown margin requirements since the design minimum boron concentrations for. assured S/D margin have conservatisms of 100 ppmB (~ 1000 pcm) built into them. Unit operations would not be ,
permissible if this procedure indicated a deviation of > 1000 pcm between calculated and predicted core reactivity, per Tech Specs.
Section 15.4.6, Chemical and Volume Control System Malfunction that Results in a decrease in Boron Con-tration in the Reactor '
Coolant, has been analyzed specifically on the basis of assured shutdown margin.
In no way does this procedure aggravate the conse-quences of this accident, on the contrary, its use J supports the mitigating assumptions made in the ;
accident analysis.
The possibility of an accident which is different ;
than any already evaluated in the FSAR will not be created by reissue of this procedure. The only '
possible accident which could result from the performance of this procedure is the failure of
- sample lines from the NC System Hot Leg. This event ;
has already been thoroughly analyzed in Section i 15.6.2, Break in Instrument Line or Other Lines from Reactor Coolant.
The probability of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased by the reissue of this proce-dure. Performance of this procedure.can in no way increase the probability of credible equipment malfunctions already analyzed per Section 15.6.2.
The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR 5_
[ .
will not be increased by reissue of this procedure.
All credible malfunctions have already been consid-ered and are bounded by the analysis of Section 15.6.2.
The pessibility of malfunctions of equipment impor-tant to safety different than any already evaluated in the FSAR will not be creet'ed by reissue of this procedure. The only active effective performance of this procedure can have is analyzed in Section i 15.6.2. Involvement with all other accidents, and the equipment malfunctions associated with them, is strictly passive in nature. These accidents are all thoroughly analyzed in Chapter 15 of the FSAR, as .
previously mentioned.-
The margin of safety in the bases of Tech Specs will not be reduced by the reissue of this procedure. -
Assurance of adequate shutdown margin as required by :
Tech Specs 3.1.1.1 and 3.1.1.2 is a primary function 1 of this procedure. Validation f fundamental as- l sumptions in the evaluation of acceptable Moderator- I Temperature Coefficient as set forth by Tech Spec 3.1.1.3 is another function of this procedure.
Reissue of the procedure in no way changes its basic-methodology and, therefore, its ability to fulfill its functions. The margin of safety assumed by the Tech Spec bases is consequently not affected by this j reissue. No USQ was-deemed to exist.
.)
MP/0/A/7150/19 Centrifugal Charging Pump 011 Cooler Removal and 1 Replacement', Change #3: This change did not involve .;
any substantive' technical or methodological revi- '
sions, but merely editorial revisions to correctly -
j indicate the intent of a step. The procedure ;
ensures that all equipment is maintained in accor- J dance with design specs and manufacturers' recommen-dations. -No USQ was deemed to exist.
.MP/0/A/7150/78 CRDM Venting Procedure, Reissue: This procedure revision is required to incorporate changes to the !
Control Rod Drive Mechatiism (CRDM) top caps per VN :
CE-1543 (Station Modification). The removal of the :
top cap eye bolts and replacement with a threaded !
plug is the only change included in this evaluation. )
3 The eye bolts installed in the CRDM top caps have {
"=
the potential to develop problems which can damage #
the CRDM vent plugs. The vent plug could be backed out or loosened during eye bolt removal which would result in NC system leakage. Westinghouse has ,
reviewed this problem and determined that the eye '
bolts can be removed and omitted or replaced with a threaded plug to prevent any potential corrosion on f
1
5 the eye bolt stud from affecting the vent plug. The eye. bolt is not part of.the NC system pressure' boundary and removal or replacement of the eye bolt will not affect NC system operability. Therefore, the probability or consequences of an accident previously evaluated in the FSAR will not be in-creased since this modification will improve vent plug reliability. The probability or consequences of a malfunction of equipment important to safety previously evaluated in-the FSAR will not be in-creased since the application of this modification will not affect the function of any equipment or systems addressed in the FSAR and will improve the reliability of the CRDM vent plugs. For the same reason, the possibility of an accident or malfunc-tion of equipment important to safety which is different than already evaluated in the FSAR will not be created, This procedure will be used for venting when re-quired and will maintain the CRDM top caps within their design requirements and specifications. No USQ was deemed to exist.
PT/1/A/4550/03C Post-Refueling Core Loading Verification, This is a Procedure Reissue: This procedure is used to verify that the core has been reloaded with the e.ssemblies in the configuration specified by the Special Nuclear Material (SNM) file core maps. Each core location is checked against the SNM maps to verify correct assembly identification and orientation and the presence or absence of an RCCA. Per the FSAR safety analysis (Section 15.4.7) this is to be done prior to each restart.
The exact method of performing the verification is not specified in the FSAR nor the applicable APM, Reg Guide, NUREG or standard. The procedure's method involves first recording the assembly iden-tification data and then independently verifying the results against the SNM maps. This method provides good assurance that any misloaded fuel will be detected. This is the' standard method use at the other Duke units.
Additional safety considerations concern the fact that the camera movements directed by this PT are classified as core alterations. This PT is per-formed under the direction of PT/1(2)/A/4150/22, Total Core Reloading. That PT contains the admin-istrative controls that verify the satisfying of the various core alteration Tech Spec requirements.
These requirements include containment isolation, source range and BDMS-alarms, manpower, minimum 2
boron concentration and level. The_ camera, its cable, and any lights weigh less'than 100 lbs total and therefore are within the bounds of the analysis in FSAR Section 15.7.4 which covers fuel handling accidents. There it is assumed that an entire assembly is dropped. The provision to allow sus-pension of ND loop operation for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is permitted by Tech Spec 3.9.8.1. It requires that full level in the cavity be maintained and that at least one loop of the ND system be operable. Thus, adequate cooling capability is maintained.
Since the methodology of this procedure is within-the assumptions made in the FSAR and provides additional protection against RCCA misplacement, the possibility or consequences of an accident previ-ously evaluated in the FSAR will not be increased nor will the possibility of an accident not already evaluated in the-FSAR be created. The only equip-ment that could possibly be affected are the RCCAs.
Since this procedure does not physically affect them but provides additional assurance that they are not misplaced, the probability or consequences of previ-ously evaluated equipment important to safety will not be increased nor will the possibility of mal-function of equipment important to safety which is different than that already analyzed be created.
This procedure is performed under the direction of PT/1(2)/A/4150/22, Total Core Reloading. That procedure ensures compliance with the various Tech Spec requirements concerning fuel handling and core alterations. No margin of safety as defined in the bases will be reduced. No USQ was deemed to exist.
HP/0/0/1004/33 Auxiliary Monitor Tank Building Vent, Original Procedure Issue: This procedure provides a method to account for activity released through the Auxil-iary Monitor Tank Building Vent. The Monitor Tank Building was installed under Nuclear Station Modi-fication CH-50180 (see 50.59 Evaluation for this NSM). This procedure will not adversely affect the-accident analyses addressed in the FSAR or present new failure modes. This procedure does not use, affect, or compromise safety related equipment. No USQ was deemed to exist.
IP/0/B/3181/24 Liquid Radweste System (WL) Containment Floor and Equipment Sump Level, Incore Instrument Sump Level, and Flow Monitoring System, Change #5: This change corrects calibration data sheet information. The floor and equipment sump level transmitters are used for rate of change of sump level calculations by the operator aid computer or control room operators if the computer is out of service. This change does 8
L not. affect the span of the transmitters; therefore,
- the calculations are unaffected, No USQ was deemed
! to exist.
T MP/0/A/7150/07 Ice Condenser Intermediate Deck Doors Testing and Corrective Maintenance: This is:a procedure reis-sue; Section ?1.3.5 was added to allow replacement of door gaskets with armaflex insulation. The door gackets are not mentioned in the FSAR and are not safety related. Station Problem' Report CNPR02029 i and Variation Notice CE-1047 were written to note the problem with obtaining spare door gaskets, and approved the use of 1-inch thick armaflex as a door gasket substitute. This change will not affect the ability of the doors to perform ttetr intended function. No USQ was deemed to exist.
MP/0/A/7150/68 Ice Basket Cable Cruciform Installation, this is a procedure rewrite; Nuclear Station Problem Reports CP-2103 (Unit 1) and CP-2146 (Unit 2) were written to allow installation of cable cruciforms into selected baskets in the ice condensers. Drawing CNM-1201.17.30/1 addresses which baskets are allowed-.
to have cable cruciforms. The changes to this-procedure clarify when the installation of ice basket cable cruciform top plates must be document-ed. No USQ was deemed to exist.
MP/0/A/7200/01 Auxiliary Feedwater Pump Turbine Governor Valve Corrective Maintenance,-this is a procedure reissue; the following changes were made:
1)_ Upgrade of Sections 2.0 through 10.0 to include more references, prerequisites, special tools, and acceptance requirements.
- 2) Disassembly of Valve, Section 11.2, rewritten to eliminate mention of "L"-Gland assembly, valve cage, and cushioning spring. These do not apply to our valve. Caution statement was added to warn against damaging the valve stem.
Addition of a dam into valve body was added to prevent debris from entering the turbine. Step 11.2.9 was added to address the removal of the carbon packing and retaining rings.
- 3) Inspection of Valve, Section 11.3, was totally.
rewritten to reflect proper valve parts, to include part number references and provide for a comprehensive inspection of the pieces subject to wear. Valve stem and bushing clearances and dimensions were included in this section.
~
- 4) Body Seat Replacement, Section 11.4, was added to allow for replacement of the welded in seat if required.
- 5) Reassembly of valve, Section 11.5, was revised to reflect changes made during disassembly of valve using the correct procedures and parts that apply. Attention was given to proper-installation of the carbon packing, washers, bushing and snap rings. Step 11.5.15 was added to blue check for 90% or greater seat contact.
- 6) Setting to Governor Valve Travel, Section 11.6, was revised to properly set-up the governor valve linkage to the remote hydraulic servo.
The manufacturers' recommendations were followed to insure full travel and control of the governor valve for all modes of turbine
< oceration. Sign-off of 5/8" travel was eliminatad because this is only an approximation. The travel is adjusted so that the valve can fully close without the chance of linkage binding. The cam plate and roller assembly is designed to give proper valve travel.
- 7) Enclosure 13.1 was revised to reflect changes in body of procedure.
- 8) Enclosure 13.2 was revised to have larger drawings for better comprehension and clarity.
These changes have been reviewed against vendor manuals, design documents, and approved station procedures. The actions outlined in this procedure' will ensure that the component is returned to as-built /as-designed conditions. No USQ was_ deemed to exist.
MP/0/A/7200/07 Turbine Driven Auxiliary Feedwater Pump Corrective Maintenance, Change #6; the following editorial changes were made:
- 1) Step 11.4.6 Deleted the words "of the first and second impeller," and added the item number of the center stage piece.
- 2) Step 11.4.7 Changed this step to make it clear that wear ring clearances on all impellers are to be checked. Also added wear ring item numbers.
m -
p <
+
- 3) Step 11.4.8'. Changed this step to make it clear that stage piece clearances are what is being checked. Wear '
ring item numbers were added.
The data sheets.were also changed-to clarify what is being checked.
These changes are not substantive in nature, and will ensure return of the pump to as-built /as- >
designed conditions. No USQ was deemed to exist..
4 MP/0/A/7600/83 Fisher Diaphragm Actuated Control -Valves Type ENA Corrective Maintenance, this is a procedure reissue: !
The Feedwater system is designed to return conden-sate to the steam generators while maintaining-proper water inventories throughout the cycle.
Feedwater is admitted to the steam generators through 4 steam generator feedwater lines, ecch of- a which contains a feedwater control valve and a '
feedwater nozzle. Feedwater flow to the individual steam generators is controlled by a feedwater control system which provides input to the feedwater '
control valves. The bypass valves around the-steam generator control valves are manually controlled-from the control room during low load conditions.
The feedwater control and control bypass valves are safety related and provide feedwater isolation to mitigate the consequences of an accident and allow safe shutdown of the reactor.
The use of live load packing in the feedwater control valves should provide better sealing capa-
'bilities to alleviate packing leakage. Design Engineering has evaluated the application of live
! load packing in the control. valves and has deter-mined 4+ not to affect valve function operation, or .i stroke time. Also, the feedwater control valves' are
,- inservice inspected in accordance with ASME Code Section XI, subsection IWV to verify operation as ,
described in.the FSAR. Therefore, the probability :
r or the consequences of an accident previously .
evaluated in the FSAR will not be increased since this procedure will increase the sealing capacity-while allowing the valve to operate as described'in the FSAR. The probability or consequences of a o
malfunction of equipment important to safety previ--
ously evaluated in the FSAR will not be increased since the application of this procedure will not affect the function of any equipment or systems addressed in the FSAR. For the same reason, the possibility of an accident or malfunction of equip- '
ment important to safety which is different than already evaluated in the FSAR will not be created.
No USQ was deemed to exist.
11
y 3
MP/0/A/7650/114 Testing of Large Bore Hydraulic Snubbers With Testan II, This is a Procedure Reissue: The ,
performance of this procedure will verify that the {
large hydraulic snubbers that protect the Steam generators are operating within specified parame-ters. No changes in as built configuration of the !
upper lateral Steam Generator support is required to -
complete the test. All testing is performed during cold shutdown (mode 5 or 6). Following testing the i snubber is returned to its original ~as built- .I condition. No USQ was deemed to exist. l 1
OP/0/A/6100/06 Reactivity Balance Calculation, Change #30; the following changes were made:
- 1) On Enclosure 4.4, change step 3.1.4 from "150-ppmb" to "170 (U1) o_t 150-(U2) ppmb".
m 2) Replace Enclosure 4.8 with new 1/M pwt. ,1 q
This change does not. affect any equipment important. I to safety except for the effect on control rods to counteract the additional reactivity of-stuck rod. j, The stuck rod boron concentration boron allowance ensures that with stuck rods the assumptions of.FSAR .;
Chapter'15.4.6 are still valid. The probability /- !
consequences of an accident analyzed in the FSAR '
will not be increased. The change (change a and b) i~ is necessary to maintain the assumptions of FSAR 15.4.6. An accident not analyzed in the FSAR will not be created. No margin of safety is affected f (other than maintaining the margin). No USQ was j deemed to exist. j a
OP/1/A/6250/08 Steam Generator Blowdown, Retype #7: This retype' includes the following changes:
L 1) previcusly approved changes 17 and 18 were incorporated. ;
- 2) Locations for manual valves were added in the j body of the procedure. l
- 3) Numerous clerical changes were incorporated.
- 4) N3M'10968 was completed. This required numer-ous changes to the procedure to provide guid-ance on operating the BB system with Cold Water Injection permanently installed.
This retype will not change the function of the BB ;
System. The changes made will result in a more efficient means of operating the system. The :
changes included in this retype will only offect the 12
non-safety related portion of the BB System located-
-in the Turbine Building.
The changes included in this retype do not involve or affect any part of a system, including the containment isolation portion of the BB System, associated _with any accident discussed in the FSAR.
The,BB System is not an initiator for any accidents previously evaluated in the FSAR or any new acci-dents; therefore, the probability' of their occur-rence will not increase due to the changes made.
The consequences of any accidents will not be increased since the changes included in this retype do not affect any_ accident mitigating system or reactor coolant system parameters. No equipment important to safety is directly_affected by the changes made.
With seismic and overpressure considered, the probabi_lity or consequences of a malfunction of-equipment important to safety previously evaluated in the FSAR will not be-increased. Changes included in this retype will not create the possibility of malfunctions of equipment important to safety different than any already evaluated in the FSAR.
Since no functional changes are being made to any safety systems'and there are no relevant parameters in the bases to any Technical Specification, the margin of safety as defined in the Bases to any Technical Specification is not reduced. No USQ was deemed to exist.
OP/1/A/6700/01- Unit One Data Book, Change #141; Excore Detector Data is being updated and the Power Range Reactor Trip setpoint is'being updated to 25% as needed for zero power physics testing.-
OP/1/A/6700/01 (Unit One Data Book) Table'2.2 is a table of data for use by plant personnel, The following describes the sections of the table and how the data is obtained and used.
FULL POWER CURRENTS This section is used to record the 100% Full Power Zero Axial offset currents and M factors for each of the Power Range Excore Detectors. Data is obtained-for thir section only by use of approved procedures such as PT/1/A/4600/05E, Refueling ENB-Calibration, or PT/1/A/4600/05A, Incore/Excore Calibration.
Other tests may supply data to this section but in all cases the tests must be approved tests.
1 The data. recorded here is used by IAE to adjust the AFD calculating circuitry and 0AC programs. It may 1
, also be used to manually calculate AFD if the OAC is inoperable.
Since AFD is used to dynamically adjust both the !
OTDT and OPDT setpoints, the data here is safety related. The accidents referenced above-depend on ,
these setpoints for mitigation. i TRIP SETPOINTS The trip setpoints for both the Intermediate (N-35 and N-36) and Power Range (N-41, 42, 43 and 44)
Detector trip setpoints and recorded on page 2 of the table. These are used by IAE.in setting the Reactor Trip setpoints on the detectors. The data .
here may be tbtained by a variety of means, i Trip setpoints for. Intermediate Range Detector may cnly be obtained by use of approved procedures to.
calculate or measure the 25% Full Power Reactor Trip setpoints.
Trip setpoints for the Power Range Detectors may not be deliberately set greater than 109% Full Power j ever. The trip setpoint may be set lower than 109% .j by use of approved procedures, by direction of Tech 3 Specs or by direction of the Shift Supervisor. The i 109% is set by Tech Spec to ensure operation is i bounded by the assumptions used in the FSAR Chapter .
15 accidents listed above. Any setpoint below 109%
may be used for conservatism or to comply with Tech i Specs.
Information in OP/1/A/6700/01 (Unit- One Data Book) is changed only by approved procedure change. It=
will not increase the probability / consequences of an- R accident analyzed in the FSAR or. create an accident not analyzed in the FSAR. No equipment other than NIS is affected by Table 2.2. The safety margin ,
will not be decreased. No USQ was deemed:to' exist.
OP/1/A/6700/01 . Unit One Data Book, Change #143; this procedure change incorporates temporary rod withdrawal limits (Data Book Curve 1.2.1) generated by PT/1/A/4150/20, Temporary Rod Withdrawal Limits Determination. The operational limitations on control rod withdrawal-allows the critical boron concentration of the Reactor Coolant System to be maintained at concen-trations low enough to ensure that the moderator temperature coef ficient (MTC) complies with Tech Spec 3.1.1.3. Imposition of these limitations is prescribed by action a. of the Tech Spec. FSAR Section 4.3.2.3 discusses behavior of MTC with 14
7 c .
i respect to NCS boron concentration, stating.that the reactivity feedback from this phenomenon becomes more negative at lower concentrations. This is the basis for the incorporation of.these-restrictions on control rod withdrawal. The rod withdrawal limits imposed by this change are at no point severe enough
-to impact the rod insertion limits specified by Tech '
Spec 3.1.'3.6, thereby having no adverse impact on assured shutdown margin (required per Tech Spec 3.1.1.1), or the core power distribution and in-serted rod worth assumed in the FSAR Chapter 15 analyses (e.g., ejected rod accident). No USQ was deemed to exist.
- 0P/1/A/6700/01 Unit One Data Book, Change #144; Table 2.2, Excore Detector Data, is being replaced. The new data reflects the resetting of power range NIS trip setpoints from 25% to 109% to allow power escalation- ,
following refueling. .The new data also incorporates.
I/R NIS 25% trip setpoint currents determined by PT/1/A/4150/21. The new data also incorporates interim P/R NIS calibration currents obtained per PT/1/A/4600/05D so that the IAE group can perform-recalibration to normalize QPTRs (which are pres-ently artificially high).
OP/1/A/6700/01 (Unit One Data Book) Table 2.2 is a table of data for use by plant personnel. The following describes the 3 sections of the table and .
how the data is.obtained and used.
FULL POWER CURRENTS This section is used to record -the 100% Full Power Zero Axial offset currents and M factors for each of the Power Range Excore Detectors. Data is obtained for this section only by use of approved procedures such as PT/1/A/4600/05E, Refueling ENB Calibration, or PT/1/A/4600/05A, Incore/Excore Calibration.
Other tests may supply data to this section but in all cases the tests must be approved tests.
The data recorded here is used by IAE to adjust the AFD calculating circuitry and 0AC programs. It may also be used to manually calculate AFD and QPTR if the OAC is inoperable.
Since AFD is used to dynamically adjust both the OTDT and OPDT setpoints, the data here is safety related. The accidents referenced above depend on these setpoints for mitigation.
i TRIP SETPOINTS The trip setpoints for both the Intermediate (N-35' and N-36) and Power Range (N-41, 42, 43 and 44) l Detector trip setpoints are recorded on page 2 of the table. These are used by IAE in setting the Reactor Trip setpoints on the detectors. The data here may be obtained by a variety of means.
Trip setpoints for Intermediate: Range Detector may only be obtained by use of approved procedures to calculate or measure the 25% Full Power Reactor Trip setpoints. These setpoints are at all times ob-tained at, and set to, power levels below the allowable limit'of 31% (as set forth in Tech Specs).
Trip setpoints for the Power Range Detectors may not be deliberately set greater than 109% Full Power ever. The trip setpoint may be set lower than 109%
by use'of approved procedures, by direction of Tech Specs or by direction of the Shift Supervisor. The 109% is set by Tech Spec to ensure operation'is bounded by the assumptions used in the FSAR Chapter 15 accidents listed above. Any setpoint below 109%
may be used for conservatism or to comply with Tech Specs.
Information in OP/1/A/6700/01 (Unit One Data Book) is changed only by approved procedure change. It will not increase the probability / consequences'of an accident analyzed in the FSAR or create an accident.
not analyzed in the FSAR. No equipment other than NIS is affected by Table 2.2. The safety margin will not be decreased. No USQ was deemed to exist.
0P/1/B/6250/04 Feedwater Heaters, Vents, Orains and Bleed Systems, Change #19; changes made to this procedure include:
- 1) Enclosure 4.18 changed to include instructions
-for Isolation and return to service of the 102 Heater Drn Tank Pump with the new mechanical seal.
- 2) New valves 1HW-297, 301, 302, 304, and 305 weret added to the valve checklist (Enclosure 4.12).
to provide initial alignment.
3)- Valves 1HW-203 and 1HW-205 were deleted per Variation Notice (VN) CE-1260.
- 4) Enclosure 4.2 changed to include instructions for supplying seal injection to the IC2 Heater Drn Tank Pump and provide guidance on how to vent the 102 Heator Drain Tank Pump with new vent valve 1HW-301.
16
IV' e .i New Mechanical seal and high point vent, installed per VN CE-2045, are the reasons for this change, e The new mechanical seal and high point vent will result in more reliable operation of pump 102. The pump will continue to function as evaluated in FSAR Section 10.4.10. This procedure change does not-increase the probability'nor' consequences of an :
accident previously evaluated in.the FSAR and it does not create the possibility of an accident not- i previously evaluated in the FSAR. '
The new mechanical seal and high point vent will not affect the function or operability of any safety related equipment. Therefore, this procedure change does not increase the probability nor consequences of. safety-related equipment malfunction previ.ously evaluated in the FSAR and it does not create the possibility of safety-related equipment malfunction ;
not previously evaluated in the FSAR.
Since no functional-changes are being made to any safety systems and there are not relevant parameters in the bases to any Technical Specification, the margin of safety as defined in the bases to any Technical Specification is not reduced. No USQ was
' deemed to exist. ,
OP/1/B/6300/01 Turbine-Generator, Change #27; changes made to this procedure include:
- 1) Added location for the generator and Exciter s Field Ground detectors (Step 1.20 Encl. 4.1).
- 2) Added additional valves ISM-21, 25, 29, 33, 155,158,-161 and 164 to be verified closed -
after the shell warming push button is pressed in step 2.9.7 of Encl. 4.1.
- 3) Note prior to step-2.9.8 of Encl. 4.1 was changed to not require computer point D2820 (Turbine Status) to indicate "NOT TRIPPED".
- 4) Added new step 2.9.9.1 to Encl. 4.1 to log the first stage shell temperature before initial pressurizing.
- 5) Added additional valves ISM-155, 158,~161 and 164 to step 2.9.14.6 and valves ISM-25 and ISM-33 to new step 2.9.14.8 to Encl. 4.1 to verify that these valves open af ter the "0FF" -
pushbutton is selected on the turbine
" Chest /Shell Warming" selector, 11 F
E 6)- Added new step 2.9.16 to Encl. 4.1 to.open valves ISM-21 and 1SM-29. These valves do not t
receive a signal to open with the off pushbut-
. ton selected.
- 7) New steps 2.15.4 and 2.22.1-to Encl.-4.1 were added to verify that valves ISM-155, 158, 161 and 164 manipulate in the appropriate positions after chest warming is selected with the
" CHEST" pushbutton or terminated with the "0FF" pushbutton.
- 8) Note prior to step.2.46 of Encl. 4.1 was deleted.
- 9) Changed step 2.49 of Encl. 4.1 from PT/1/A/4250/02A to PT/1/B/4250/02A.
- 10) With 3-Arc admission, new steps 2.1.1.6.1, 2.1.1.6;2, 2.2.1.5 and 2.2.1.6 of Encl. 4,2 were added to verify that valves ISM-25 and ISM-33 manipulate with the-opening and closing.
of their respective control valve.
- 11) New step 2.3.2 was added to Encl. 4.4 to veri _fy that valve 1SM-25 opens when CV3 fully closes (approximately 65% of full load).
- 12) Restricted change #26_was-deleted.
Changes #2, 3, 5, 6, 7,-10, 11 and 12 are being made because of the implementation of Nuclear Station Modification CN-10779. Change #8 is being made because of the implementation of Nuclear Station Modification CN-10668. Changes #1, 4, and 9 are being made to enhance the effectiveness-of this procedure.
Modifications made in Nuclear Station Modification CN-10779 and Nuclear Station Modification CN-10668 were evaluated in their_ appropriate Safety Evalua-tion and found to not contain any unreviewed safety questions. Because the majority of the changes made-in the procedure change result from the implementa-tion =of the previously. named NSM's, these changes do not contain any unreviewed safety questions.
Changes #1, 4 and 9 are being made to improve the ease and effectiveness of this procedure and do not create any unreviewed safety questions.
Since no functional changes are being made to any Safety. Systems and there are no relevant parameters in the bases to any Technical Specification, the
- margin of safety as defined in the bases to any M
o; ,
Technical. Specification is not reduced. No USQ was
< deemed to exist, OP/2/B/6400/01B Condenser Tube Cleaning System, Changes #7 and #8:
This' procedure change is to replace the 33mm amertap balls with size 32mm amertap balls in the main condenser. These new 32mm balls will be used'in the main condenser until the Unit 2 refueling outage.
At that time the condenser tubes will be inspected to determine the performance of the 32mm balls.
The supplier of the amertap system has recommended-the use of 32mm hard balls instead of 33mm soft balls for good scrubbing and longer life. This change does not involve or affect any part of a system associated with any accident discussed in the #
FSAR. The RA system is not an initiator for any cccidents previously evaluated in the FSAR or any new accidents; therefore, the probability of their occurrence will not increase due to this change, The consequences of any accidents will not be increased since this change does not affect any accident mitigating system or reactor coolant system parameters. No equipment important to safety is directly affected by this change.
Since no functional changes are being made to any safety systems and there are no relevant parameters in the bases to any Tech Spec, the margin of safety as defined in the bases to any Technical Specifica-tion is not reduced. No USQ was deemed to exist.
PT/0/A/4150/20C Westinghouse Digital Reactivity Computer Checkout, this is an original procedure issue: :This PT is used to perform three functions. .The first is to check the calibration of the analog to digital'and digital to analog converters on the Westinghouse Digital Reactivity Computer system, The calibration method is in accordance with the recommendations supplied in the vendor documentation. The second function is to check that the physics constants are correctly input into the software. This is done by comparing the values input into the computer with the Design Engineering values, The allowance of a one unit difference between the two sets of values for the Total Beta Fraction and Beta Effective (Acceptance Criterion 11.2.2) is required to account for roundoff errors in the Design Engineering values. The third function is to test the system (hardware and software) by performing measurements using exponential test signals for input. This exponential testing is equivalent to that recom-mended by Westinghouse for calibration of the analog reactimeter. These calibrations qualify the 19
___.__MML_2.'._;_.~'..'__.' .
t (* /
computer to be used during Zero Power Physics Testing as the means for measuring reactivity during-various tests. These tests involve the measurement 6
of control rod worth, critical boron . concentration, and moderator temperature coefficient.. However, before these tests,-an additional functional check-is made of the reactivity computer .using the. reactor itself as the signal source. This check is per-formed as part of the controlling-procedure for-physics-testing. .This provides additional verifi-cation of the validity of the reactivity results.
This PT does not directly affect any equipment-important to safety. It is used to calibrate instrumentation used to verify design predictions of important physics parameters. These parameters include control rod worth, critical boron concen-tration, and moderator temperature coefficient.
Verification of the design predictions assures that the physics assumptions made in numerous FSAR accident analyses are not violated. This PT in-creases assurance that these measurements are accurate. Therefore, the probability or conse-quences of an accident or malfunction'of equipment important to safety will not be incr' eased. The margin of safety in the Tech Specs will not be-reduced. No USQ was deemed to exist.
PT/1/A/4150/05 Core Power Distribution, this is a procedure reis-sue; the following changes have been made to this procedure:
- 1) The method of acquiring reactor coolant flow.
data during flux mapping is enhanced. The new method involves trending and averaging of this-data for the duration of the flux map as opposed to the old methoc' of obtaining a single
" snapshot" of this data. This periodic test's surveillance of NC Flow with respect to "R"'
(which is calculated from the Enthalphy Rise Hot Channel Factor, FAH, obtained by the flux-map) is the means by which proper margin to core DNBR is verified. The basis for DNBR is discussed in FSAR Section 4.4.1.1 and the relationship of NC Flow and "R" is analyzed by Tech Spec 3/4.2.3.
- 2) The method of acquiring .Incore Thermocouple data has been expanded to a program of trending and averaging over the duration of' flux map-ping. The old method entailed " snapshots" of T/C indications before and after the flux map.
With this new method, more representative T/C indications are obtained for the purpose of 20
J 1
calculating T/C Mixing Factors, which are .
correction factors generated on the basis of ,
the power distribution (FAH's) measured by the 3 flux map. Requirements for.the calculation of '
the Mixing Factors and their incorporation into the OAC are also added to this periodic test as n part of the~ reissue. With new T/C Mixing t Factors generated from each flux map obtained, the validity of the Incore T/C power distribu . ,
. tion (FAH's) calculated by 0AC Program NUCLEAR' 11, Thermocouple Map, will be optimized, . The !
basis.for a reliable core power distribution
.from the Incore T/C's is stated in FSAR Section 4.4.6.1 which refers to the Incore T/C's as a backup for the Incore Flux-Mapping system.
- 3) Detailed guidance-for executing the SNC Core Program to analyze the raw incore detector data r; obtained per the flux map has been added. }
Previously, little guidance was- provided-for this. . This change allows relatively inexperi- ,
enced. personnel to quickly and accurately 3 execute this program to obtain the analysis required to meet the surveillance of the core peaking parameters described in FSAR Sections 4.3.2.2.1 through 4.3.2.2.6. !
- 4) A step for notifying Operations personnel of any flux map findings in violation of Tech Spec '
limits or. requiring restrictions on unit operation has been added. This change will "
ensure that Operations is fully appraised of the unit's measured core power distribution with respect to operating limits set'forth by ,
the FSAR and Technical Specifications so that ;
appropriate actions are taken when required.
Neither the probability nor the consequences of any previously analyzed accident will be increased by this procedure reissue. The aforementioned changes -
enhance the ability of the procedure to accurately monitor core power' distribution, thus assuring compliance with' Tech Specs (which ensures that core power distribution is within assumptions of FSAR analyses). Acceptable margin to DNBR is thereby verified. As stated in FSAR.Section 4.4.1.1, com-pliance with DNBR limits is an integral part of the design basis for the spectrum of Condition II events described in Section 15.0.1.2 of the FSAR. These changes have no affect on analyses of Condition III and IV events described in Chapter 15.
The possibility of an unanalyzed accident or safety significant equipment malfunction as a result of 21 1
+.
?
? l r
i this reissue will not be increased. The aforemen-
,4 tioned' changes are passive in nature, having no.
direct impact on unit operation. The introduction of erroneous Incore T/C Mixing Factors would have no I affect on Operator response since the power distri-bution inferred by them is not used in any way for unit (or equipment) operation.
Neither the probability nor the consequences of any previously analyzed equipment malfunction with safety significance will be increased by this reissue. The only plant equipment impacted by this change are the Incore Thermocouples (described in FSAR Section 7.7.1.9.1). As previously mentioned, no error made with respect to generation and incorporation of correction factors for these thermocouples will impact the reliability of these components. This is due to the fact that the raw temperature (in F)indicationoftheT/C's,not their inferred power distribution, has safety -
significance. These temperature indications are j unaffected by this change. -
The margin of safety will not be reduced for any of d the applicable Tech Spec bases by this reissue. On j the contrary, all safety margins will be assured by j measures taken by the changes made by this reissue 1 to enhance the surveillance capabilities of this periodic test. No USQ was deemed to exist. i 3
PT/1/A/4150/10 Boron Endpoint Measurement, Change #10; there are ->
now three. types of reactivity computers available.
Part of this change adds the third type (Westing- '
house digital reactivity computer) to the choice of ,
tests required to be performed as a prerequisite. l The Westinghouse digital reactivity computer is j tested to the same or better accuracy requirements q as the other two types of computers that could be- :
used for this test. The Westinghouse digital .
reactivity computer is tested per PT/0/A/4150/20C, I Westinghouse digital reactivity computer checkout. !
The software is controlled per Station Directive 2.0.8 (handling of "Off-line" computer programs and' calculationmethods).
Catawba is committed (to the NRC) to the Duke Power Company McGuire/ Catawba Nuclear Station Startup Test-Program dated April 1988, CN-524.00. This change incorporates the initial conditions for this par-ticular test from the startup test program into the body of the procedure. No USQ was deemed to exist.
PT/1/A/4150/118 Control Rod Worth Measurement by Rod Swap, Change
- 8: This change is to provide a means of completing 22
k the Rod Swap method of rod worth measurement when
.the initial rod configuration is not the one assumed by the PT. The assumed rod configuration is ARO except for the Reference Bank (RF) nearly fully inserted. However, the' initial configuration for this performance of this test was AR0, bank RF fully inserted, and bank 1 (first bank to be swapped) slightly inserted. This configuration resulted in the need to modify the calculations in the PT to correct for the bank 1 insertion. This ensures that-the results of the test are accurate and that the DDE-MN predictions of rod worth are validated. This in turn validates the core design and supports the conclusion that the core design is within the FSAR assumptions for such parameters as rod worth inser-
~
tion, shutdown margin, And stuck / ejected rod worth.
The other potential impact on nuclear safety is that the initial rod configuration had more rods inserted than assumed by the PT. This means that the shut-down margin could be less than assumed. The Con-trolling Procedure for Startup Testing (PT/1/A/4150/21) contains an Enclosure (13.4) to verify that shutdown margin is maintained during the rod swap PT, It shows that the shutdown margin during rod swap assuming only bank RF fully inserted is about 4500 pcm (including several margins for conservatism). Note that Tech Specs require mini. mum of 1300 pcm. Even if bank 1 had also been fully inserted, the shutdown margin would be about 4000 pcm. This is based on a predicted worth of bank 1 of 300 pcm (individual worth) and less than 600 pcm (worth in combination with other inserted banks).
This' combined worth is based on historical values of alpha (ratio of worth of bank individually vs.
combined). It is typically a value no more than 1.3. The difference between rod worths measured individually'and combined can also be estimated by the ratio between the total of the individual worths-(4974'pcm) and the total combined worth (7553 pcm).
This ratio is about 1.5. The maintaining of ade-quate shutdown margin is an important assumption made in many FSAR analyses.
This change does not increase the probability or consequences of an accident analyzed in the FSAR.'
The possibility of an accident not previously evaluated in the FSAR is not created. No equipment important to safety is affected by this change. The a rgin of safety as defined in the Tech Spec bases is not reduced No USQ was deemed to exist.
PT/1/A/4150/19 1/M Approach to Criticality, this is a procedure reissue; the following changes were made:
2_3 i
, 1) In Part 1, Criticality Following Refueling,'it adds guidance for the prediction of ICRR behavior during dilution of the NCS to the desired critical boron concentration prior to control rod withdrawal. Presently an arbitrary ICRR value of 0.3 is identified as a threshold for concern if extrapolation of the ICRR during dilution forecasts that ICRR will be below this value when the designated quantity of water -is added to achieve dilution. Extrapolation to below ICRR = 0.3 would shed considerable doubt on the validity of predicted reactivity param-eters of the core design and necessitate a halt of dilution upon reaching an ICRR of 0.5-pending evaluation of design data. The value of 0.3 is arbitrarily low, however, so a method of deriving a more realistic projection of ICRR behavior has been introduced. This method incorporates-the initial NCS boron concentra-tion for criticality and the cumulative HZP Shutdown and Control Rod Banks worth to esti-mate ICRR behavior during dilution. The ICRR projection obtained by this method is adjusted downward by 10% to account for reasonable uncertainties in the predicted rcactivity data.
The adjusted projection is plotted on the graph of "ICRR vs. Demin Water Added" for comparison against the actual ICRR obtained during dilu-tion. A restriction for halting dilution when ICRR reaches any point on projection line if extrapolation predicts the ICRR being below the ICRR projection at the conclusion.of dilution, replaces the restriction to halt at an ICRR of-0.5. This methodology employs conservatism with regards to Nuclear Safety in that it, 1)
Projects ICRR behavior based on a NCS boron concentration which assures required shutdown margin,-2) Carefully monitors actual ICRR with respect to " projection" as the BDMS is period-ically reset during dilution, 3) Flags prema-ture approach to criticality based on compari-son to projected ICRR and requires halt of.
dilution. This method is therefore.a backup to the BDMS during the dilution phase of approach to criticality. This change will allow early notification of potentially significant errors in design data.
- 2) Guidelines for the removal of.the spare scaler / timer (installed on one of the Source Range NIS Channels prior to core reloading) have been added. This action is to be per-formed in Mode 2, following criticality, so no Tech Spec requirements for the Source Range NIS 2_d
--mm_m--,-- -,.~-.,_a.u-w--n,--,.,:.-,.---
-- a ..u- - - ~ . . - __-..-,- u---_-.w.~. --- .a.--, . < . - - - ---..c. .- - --.s - - - a- - ~ . -- -- c .-.-m-,--,-- . - , _u--.u.------u,u---,a.-.-.-------,,,-,.u.__
i
'I are affected. These instructions have been i developed carefully to ensure operability of !
i the NIS is not degraded (for subsequent utili- l ty, as stated in Tech Spec 3.3.3.12) in any '
way. '
- 3) Complete revision of the method by which _
criticality is achieved in the event that the' predicted Estimated Critical Rod Position (ECP) !
is missed during routine unit startup-(Part 2 of this procedure) is incorporated by this' -j reissue. The present method has no guidance- i for the determination of an Estimated Critical i Boron Concentration (ECB) at which criticality l could reasonably be expected to be achieved Lj during a subsequent effort. This reissue introduces formal guidance for this determina-tion as follows:
y
- The ICRR plotted by Operations during the '
aborted startup attempt is obtained by the Test Coordinator (TC).
- The TC extrapolates from the last two l points plotted on the ICRR to obtain a i Predicted Critical Position (PCP).
- Inferring the equivalent reactivity worth of the span between ECP and PCP'and accTinting for transient xenon behavior prcjected to occur by the time of the next ;
startup attempt, the TC' calculates a new ECB, ,
i
- Operations then calculates a new ECP and '
reattempts criticality as the TC performs an ICRR.
This method has the advantage of providing an educated estimate of CCB for the subsequent approach to criticality. It incorporates all the caution of any other routine approach to i criticality during cycle operation since an i ICRR is performed by the TC during the repeat startup as part of this procedure. It thereby complies fully with all features of " Dilution During Startup (Mode 2)" discussed in FSAR Section 15.4.6.
25
g ,
Neither the probability.nor the consequences of ,
any accident previously evaluated in the FSAR will be increased by this procedure reissue.
All control rod manipulations are performed '
under approved operating procedures during-approach to criticality. The analyses of-FSAR Sections 15.4.1, 15.4;2, 15.4.3, and 15.4.8; addressing various accidents propagated by RCCA~ .
malfunctions, therefore bound all such manipu- '
lations since they acccunt for'.such normal operation of the RCCAs.
Adjustment of the NCS. boron concentration directed by this procedure is likewise per- ;
formed in accordance with approved operating ,
procedures. This action is therefore bounded' by the analysis of FSAR Section 15.4.6 dealing with CVCS malfunctions resulting in decreased 1 reactor coolant boron concentration. .,
Neither the probability nor the consequences of any safety significant equipment malfunction- '
previously ar.alyzed in the FSAR will be in-creased by this reissue. The plant components-specifically involved with this procedure are the control rods, the CVCS's reactor makeup '
subsystem, and the Source Range NIS, No abnormal' demands are placed upon any of this equipment thereby prec1'uding unanalyzed mal-functions. ,
The possibility of-an accident or safety significant equipment malfunction not'previ-ously analyzed in the FSAR'will not be created by this reissue. This is due to the fact that the unit is not placed in any unanalyzed configuration nor is any safety related compo-nent operated in an abnormal mode by the changes made by the reissue. All potential events are therefore fully analyzed.
The margin of safety defined in the. bases of Tech Specs impacted by this procedure are not
reduced in any way. The subject safety margins are assured procedurally by such specific means is:
- Limit and Precaution to assure T-avg as ;
maintained > 551 F in Mode 2 (Tech Spec '
3.1.1.4).
- Limit and Precaution and procedure steps in place to assure adequate Shutdown Margin (Tech Spec 3.1.1.1).
L6
p..
- Procedure steps to avoid criticality below Rod Insertion Limits (Tech Spec 3.1.3.6).
- - Procedural assurance of Source Range NIS operability (Tech Spec 3.3.3.12).
No USQ was deemed to exist.
PT/1/A/4150/21 Post-Refueling Controlling Procedure for Startup -
Testing, Change #12: This change allows an alter-
@ nate method of calculating the top of the zero power e physics test band. The change in method does not reduce the 50% minimum margin required between the upper limit of the test band and the point of adding nuclear heat. It is important that this margin be 3 maintained so that the effects of nuclear heat are ;
4 prevented from adversely affecting reactivity measurements. These measurements are used to verify design engineering predictions of certain' physics parameters (e.g., RCCA Bank Worth, ITC, etc.) This change is in accordance with the " Duke Power Company McGuire/ Catawba Nuclear Station Startup Physics Test Program".
All testing is performed at zero power. P/R flux 1 setpoints are set down to 25% during.this time. ;
There ere no accidents which are affected by the .;
point determined to be. set as the upper limit of the zero power physics test band. No USQ was deemed to exist.
PT/1/A/4200/01F Lower Personnel Air Lock Leak Rate Test, Change #28:
This change is restricted to two days of testing (1-18-89 to 1-19-89), and allows the use of two strongbacks in place on the lower airlock door instead of the normal three due to an obstruction t which makes it impossible to_ install three. Two +-
strongbacks have been used in the past successfully, j Since containment integrity and thus the airlocks J are not required to be operable'in Mode 5, no ,
postulated accident scenarios are affected. The strongbacks are used as an additional safety pre-caution and past experience has shown that the test can be safely performed without the third strongback, t No USQ was deemed to exist.
PT/1/A/4200/01N Reactor Coolant System Pressure Boundary Valve Leak l Rate Test, this is a procedure reissue to incorpo- 1 rate previous changes into a retyped procedure, in ,
addition to the following changes: (
- 1) Format of enclosures for valves tested on Enclosures 13.1 through 13.26 has changed.
21
l
- 2) Enclosures 13.27 through 13.29 added to allow !
testing of groups of valves, i
- 3) On Enclosures 13.30 and 13.31, test method to-be used if Residual Heat Removal Valves IND2A -
(IND37A) leaks greater than IND1B (IND36B) has changed. ,
i PT/1/A/4200/01N is written to satisfy the Surveil-lance requirements of Tech. Specs. 4.4.6.2.2. The testing is done to ensure early detection of possi-ble in-series check valve failure. The valves which are tested by this procedure are important in preventing overpressurization and rupture of the ECCS low pressure piping which could result in a LOCA that bypasses containment.
During the test, the valve under test is pressurized so that the differential pressure across the valve is at least 100 psid. Valves are pressurized in one of two ways: either directly by NC System pressure (for first stage check valves and for IND1B and IND36B) or by NC System prenure after opening the bypass line around the first stage valve (for second stage check valves and for IND2A and 1ND37A), When
~
testing the pum) suction valves, a determination is made as to whic1 of the two valves in the ND Pump suction line is leaking the greatest. The leakage for the valve leaking the greatest is then measured and is assigned to both valves in the suction line. ;
Neither the probability nor the consequences of an ,
accident will be increased by this test. This test '
is perfornied to ensure low probability of gross failure of the check valves. During the test, the ECCS low pressure piping is isolated; therefore, overpressurization of the ECCS low pressure piping will not occur while one of the two in-series check valves is being tested if the check valve being tested were to fail. Neither the probability nor the consequences of a malfunction of equipment important to safety will be increased by this test.
This test is performed to verify the integrity of the valves, which are important in preventing overpressurization and rupture of the ECCS low ,
pressure piping which could result in a LOCA that bypasses containment. During this test the low pressure. piping is isolated and the testing is done at a pressure lower than the pressure at which the .'
pipe is rated. The margin of safety as defined in the bases to Technical Specifications will not be reduced by the performance of this test. No VSQ was deemed to exist.
[ ;
l
-PT/1/A/4200/09 Engineered Safety Features Actuation Periodic Test, !
Change #69: The first two items of the change {
p^ simply return test switches to their normal posi- ;
tion.- The steps were inadvertently left'out of an i
earlier change. The third item of the change ;
deletes the retest of CRDM Vent Fan 10. The fan ;
L motor will not be replaced this outage; therefore, ,
I the fan cannot be retested. The fan is normally tested to ensure that the D/G load sequenced will '
c load the fan onto the D/G during a blackout.
g Neither the probability nor the consequences of any accident will be increased by this change. The :
Control Rod Drive Mechanism (CRDM) Ventilation ;
g System consists of four 33 1/3 percent capacity fans '
and associated dampers and ductwork. Three fans L normally operate. The containment ventilation P system is not an engineered safety feature and no l credit has been taken for the operation of any .
subsystem or component in analyzing the consequences J of any accident. Neither the probability nor the consequences of a malfunction of equipment important to safety will be increased by this change. CRDM Fan 1C will be remeved from service and the three remaining fans can supply the necessary cooling requirements. The margin of safety as defined in the bases to any Technical Specification will not be reduced. No USQ was deemed to exist.
PT/1/0/4250/12 Cendenser Tube Leak Detection. System Operability-Test, this is an original procedure issue: This procedure is used to periodically test and document the operability of the condenser tube leak detection system. It is an administrative procedure to ensure adequate performance during periods of condenser tube leaks. Equipment operation is' conducted per OP/0/0/6250/13, Operating P,ocedure for the Con-o denser Tube Leak Detection System, which is a previously evaluated and approved procedure. No safety related equipment is involved in this proce-dure. No USQ was deemed to exist.
PT/1/A/4350/15A Diesel Generator 1A Periodic Test, Change #17:
Nuclear Station Modification CN-11104 replaces D/G 1A.10-10 lube oil and overspeed emergency trips.
(pneumatic) with electronic trips. During perfor- .f mance of D/G 1B retest, problems were identified and resolved. The test method was changed to include the new solenoid and change of contact status. This change makes the D/G 1A periodic test the same as j D/G 18 periodic test which has already been com- i pleted. D/G 1A will be inoperable at all times during performance of this test. Only after com-pletion of this test and the operability test will D/G 1A be declared operable. This change will !
I 29
verify the electronic trip signal, which is a more '
reliable trip signal. No USQ was deemed to exist.
PT/1/A/4350/02E CA (Auxiliary Feedwater), CF (Main Feedwater), and Turbine Interlocks Periodic Test, Change #39: The purpose of this change is to assure that the cir- '
cuits being tested are not affected by the Antici- '
pated Transient Without Scram (ATWS) circuitry that was installed by Nuclear Station Modification (NSM) !
CN-10952. .
No components will be tested differently as a result of this change. This change merely isolates the ATWS inputs fres '.he normal circuitry that is being ;
verified. Although ATWS will be effectively dis-abled, it is not required in the modes that the test may be performed in. Therefore no accident scenar- c ios are affected in er . " and no new accidents !
scenarios are create... The quipmer.t operation is not affected in any way, so ,either the probability 3 of nor possibility t i any ,.alfunction of equipment important to safety a e , creased. The margin of i safety as defined in t" bases of Tech. Specs, is not reduced since the tost method and resulting :
actions are not changed. No USQ was deemed to i exist. ,
PT/0/A/4400/01A Exterior Fire Protection Functional Capability Test, this is a procedure reissue: This procedure is rewritten to meet requirements of National Fire r Protection Association Section 20 (NFPA-20) re-quirements. NFPA-20 requires three pump head points -
to be obtained at:
f
- 1) Pump design point 2500 gpm at Total Head of 144 psid. ;
- 2) 150% Pump Flow 3750 gpm at 65% of Total Head '7 93.6 psid. ;
i
- 3) Minimum Pump flow which is 1000 gpm.
This third point cannot be met due to design deft-ciencies, Instead an arbitrary third point will be ,
obtained until the design deficiencies are resolved. t Limit and Precaution 6.7 addresses overpressuriza- 1 tion concerns.
This procedure will align one RY pump to pump through the RY Test Header which makes the pump inoperable. Only 2 out of 3 pumps are only required to be operable at all times, and 1 out of 3 pu:nps ;
operable requires action within 7 days. Therefore, Limit and Precaution 6.3 addresses the requirement ,
30
?
.1
l s of at least one operable pump. No USQ was deemed to ,
exist.
PT/1/A/4450/03C Annulus Ventilation System Performance Test, Change
- 6: This is a restricted change, applicable between the dates 1/17/89 to 2/10/89.
The Annulus Ventilation System is required to be l OPERABLE in Modes 1, 2, 3, and 4. A possible ,
challenge to operability has been identified by '
performance involving the performance of t PT/1/A/4200/01F, Lower Personnel Airlock Test.
During one section of the airlock test, it is .
necessary to remove a 2 foot square plate in the annulus boundary. This test will establish whether the Annulus Ventilation System can still meet Technical Specification 4.6.1.8.d.4 with the annulus boundary plate removed:
"4.6.1.8.d.4 Verifying that each system proriuces :
a negative pressure of greater than i or equal to 0.5 inch Water Gauge in the annulus within 1 minute after a start signal."
- Because an operability question exists, the modifi-cation (removal of the plate in the annulus bounda- .
ry) will be made, the Annulus Drawdown Test will be !
performed to verify Tech Spec 4.6.1.8 d 4, and the modification will be restored to normal condition '
prior to Mode 4 entry. ,
Since this experiment is being performed only in modes other than 1, 2, 3 or 4 (modes in which VE is required to be OPERABLE), the probability of an :
accident previously evaluated in the FSAR will not be increased. The consequences of an accident previously evaluated in the FSAR will not be in-creased. The possibility of an accident which is '
different than any already evaluated in the FSAR :
will not be created. The probability of a malfunc-tion of equipment important to safety previously evaluated in the FSAR will not be increased. The consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. The possibility of malfunctions of ,
equipment important to safety different than any already evaluated in the FSAR will not be created.
And the margin of safety as defined in the bases to any Technical Specificatien will not be reduced. No USQ was deemed to exist.
PT/1/A/4600/05E Refueling ENB Calibration, Change #3: This proce-dure calculates preliminary setpoints for the rod 31
?
t i stop and reactor trip functions of the Intermediate '
Range (I/R) NIS, prior to startup following refuel-ing, based upon the core power distribution measured near the end of the previous fuel cycle and the core power distribution predicted for the beginning of -
D the new fuel cycle. These setpoints (corresponding ,
c to 20% F.P. and 25% F.P. respectively) are derived ;
by adjustment of the old setpoints with a ratio of t the measured and predicted core power distribution data. This change allows I&E to waive the adjust- +
ment of the I/R NIS setpoints-if the calibration currents provided by the procedure correspond to :
setpoints which are within 3% (of full power), on i the high (conservative) side, of the present setpoints. This is permissible with regards to the i nuclear safety functions of these setpoints since L they would be impacted at a lower reactor power -
level. This conservatism ensures that no safety analyses are affected by this change. Adjustment of these setpoints would be performed in all cases in which the new calibration currents are lower (and therefore non-conservative) than the present 20%
F.P. and 25% F.P. setpoints. Conservative actuation '
of the I/R NIS rod stop and reactor trip setpoint safety functions is not degraded by this change.
Allowing the actual I/R reactor trip setpoint to be inaccurate up to 3% in the conservative direction (i.e., 22% F.P.) will not result in an unanticipated-reactor trip when the I/R trip function is enabled by P-10 (Power Range NIS $ 10% F.P.) during power reductions. This assessment is based on an evalua-
tion of the expected changes in I/R and P/R NIS response due to heavy control rod insertion during power reductions. This evaluation concludes-that it is not possibic for the indicated power of the I/R .
NIS to be 3 22% F.P. at a point in which P/R NIS is indicating 5 10% F.P. No USQ was deemed to exist. q PT/1/A/4600/05F Running Incore/Excore Calibration, this is a proce-dure reissue; changes include:
- 1) Made editorial changes to " PURPOSE".
- 2) Added references 2.4, Tech Spec 4.3.3.2 on 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> calibration of Incore Dete: tors, and 2.5, Memo to File regarding RPECALIB program.
- 3) Deleted Prerequisite' Test 4.4, PT/1/A/4150/03A, !
p NSSS Thermal Outputs.- PT/1/A/4150/03A is not required to be completed before starting this L test per pT/1/A/4150/21, Post Refueling Con-trolling Procedure for Startup Testing. i PT/1/A/4150/03A is performed to verify 0AC program is calculating thermal power correctly 3.2 1
O for a specific set of inputs. In the unlikely case that a problem exists with the OAC pro-gram, the problem would be detected and re-solved before increasing power above the 80%
plateau. Effects on this calibration would be evaluated and corrections m6de, if necessary.
- 4) Added Limit and Precautions:
a) 6.13 to warn of the possibility of re-ceiving a rate trip on a power range channel while obtaining currents from that channel, b) 6.14 to describe the undesirable effect of excessive control rod insertion on-cali-bration, c) 6.15 to specify minimum acceptablo Axial Offset span of 10%.
- 5) Changed general sequence of test to take Incore and Excore data between 65% and 80% power, inclusive, beginning essentially All Rods Out at 65% and driving rods to obtain an AFD change of approximately -2% for each 2.5% power increase. Old sequence was to take data' between 50% and 80%, with AFD change ~of ap-proximately -1% for each 2.5% power increase.
New sequence will result in higher average power for test, reducing extrapolation errors, and potentially higher average rod insertion, reducing error due to rod insertion. The old sequence is known to produce invalid results (in 3 out of 3 previous attempts); the new sequence attempts to correct the sources of error in the old sequence; The number of flux maps has-been reduced from 12 to 7 because of the reduced power span for map. Review of results from some past incore/Excore Calibra-tions indicate that 5 maps are sufficient to determine the line of current versus Axial-Offset.
- 6) Change Prerequisite system condition 8.4 to specify that Control Bank D is 1 220 steps withdrawn and deleted not allowing step to be-waived, due to the concern over the effect of rod insertion on excore currents.
- 7) Deleted from Data Required Thermocouple Maps and Nuclear 06 printouts. This data was not used in procedure.
i x _
r I
i, L
L
- 8) Added " bang-bang" maneuver to stop axial Xenon oscillation and return AFD to target when 80%
power is reached. Maneuver is estimated to save 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when waiting for sufficient stability to perform Core Power Distribution at 80%.
- 9) Deleted Acceptance Criteria regarding indicated excore tilt. It is not the purpose of this procedure to perform the Tech Spec Surveillance of QPTR. Calibration data is checked to verify tilts are " zeroed" out as a result by using
, currents obtained with last flux map.
p 10) Added Step / Enclosure to create a theoretical factor file for analyzing the Quarter Core Flux.
^
Maps taken during this procedure.
- 11) Added Enclosure to provide guidance on analyz-ing the flux maps.
- 12) Changed name of procedure to be more concise and consistent with other procedures used to develop calibration data for power ranges.
The purpose of PT/1/A/4600/06F, Running !
Incore/Excore Calibration, is to collect incore flux map data and excore power range data at various Axial Offsets while increasing power during initial ,
power escalation following a refueling outage. Data is used to calibrate power range channels so that excore Axial Flux Difference (AFD) is consistent ,
with incore measurements. Another result of cali- t bration is to normalize power range tilt ratios to !
approximately 1.00.
The procedure method has been extensively modified [
due to the poor correlation of excore AFD to incore measurements experienced following previous perfor-
- mances of this procedure. The inaccuracies have been attributed to excessive contrcl rod insertion, which decreases the flux seen by the power range detectors, and extrapolation of currents from relatively low power to full power.
Excore AFD and tilt ratios are the primary real-time .
indications of Core Power Distribution. As such, there are Technical Specifications associated with .
both AFD and Quadrant Power Tilt Ratio (QPTR), which ,
is defined as the maximum power range tilt ratio.
Operation within the Tech Spec Limits on AFD and
, QPTR ensures that the core power distribution ,
between monthly incore flux maps remains within the o power distribution assumed in the FSAR accident 1
s +
g analyses. A function of AFD, f(AI), is included as a term in the over temperature delta temperature (OTAT) trip setpoint formulation. The f(AI) func-tion reduces the OTAT trip setpoint for axial power distributions that are more positive or more nega-tive than a specified dead band.
This procedure is written to ensure that the re-quired data is obtained safely and efficiently. AFD is driven negative by control rod insertion as power is increased from 65% to 80% power at 2.5% per hour.
During test, rod insertion limits, withdrawal limits (if applicable), and AFD limits per Technical Specifications will be observed at all times.
Anticipated AFD change will be 12%. Measurements over a wide range of axial power distributions are necessary in order to determine a relationship between incore and excore measurements that will remain valid at extreme axial power distributions.
A number of steps are taken to ensure that calibra-tion data is accurate. Incore and excore data are analyzed using approved and controlled software. A least squares fit of the data is performed for each power range detector, and a ' fit correlation and error for each point used in the fit are calculated so that bad data points may be excluded from the fit. Calibration data is used to calculate excore AFD from the last map and compared to incore results to ensure consistency. Tilt ratios are also calcu-lated in the same manner and verified to be approx-imately 1.00. Af ter IAE performs the power range calibrations and calibration data entered as re-quired for the OAC Excore Power Distribution Moni- i tor,-power range currents are obtained with OAC data. AFD and tilt ratios from hand calculations are compared to OAC indications to verify consis-tency. i The sest does not require any unanalyzed operation - ,
with respect to core power aistribution. Systems i used to support' test, such as the rod control system ,
and the chemical and volume control system, are used '
in normal operational modes per approved operating procedures. The moveable incore detector system is also used in accordance with an approved operating procedure. Excore detector data is obidned without any abnormal configuration of power range channels; they remain fully functional throughout the data gathering portion of the test. Obtaining current data from a power range has infrequently resulted in receiving a rate trip on that channel. This has no effect on the Unit unless another channel is tripped, in which case a Reactor Trip would occur.
L5
i A limit and precaution is included to remind the user of this possibility and to specify that the c user should verify that all channels are in normal l status (i.e. ILot tripped) before opening a power range drawer. If a Reactor Trip were to occur, the transient would be bounded by analyzed transients.
, Calibrations of power range. channels are done in t
accordance with approved instrumentation procedures which require that the channel being calibrated be removed from service and associated bistables tripped. Reactor trip logic is thereby reduced to 1/3 logic, which is conservative to the normal 2/4 logic, t
For the above reasons, it is concluded that the probability or consequences of an accident evaluated in the FSAR will ILot be increased, nor will the i possibility of an accident not evaluated in the FSAR :
be increased. Likewise, the probability or consequences -
of a malfunction of equipment important to safety ,
evaluated in the FSAR will not be increased, and the possibility of an unevaluated malfunction will ,n.ot be increased. !
As operation will remain within Technical Specifi- .
r cation limits at til times during the test, the ,
margin of safety will not be reduced. '
The FSAR test abstract for Incore and Nuclear ,
Instrumentation Systems Detector Correlation con-tained in Table'14.2.12-2, Initial Startup Testing, provides a method for performing power range cali- ,
bration. This procedure uses a different method for performing the calibration. Abstracts contained in i; Table 14.2.12-2 are not applicable to surveillance tests such as this one (except as used for initial startup testing). Therefore, no change to the FSAR -
a abstract is required. No USQ wcs deemed to exist. l PT/2/A/4200/01A Containment Integrated Leak Rate Test, this is an .
original procedurc issue'. PT/2/A/4200/01A, Con- !
tainment Integrated Leak Rate Test is performed in ;
modes 5, 6 or No Mode to satisfy Surveillance t Requirements 4.6.1.2, 4.6.1.6 and 4.6,1.7. The test !
is outlined in FSAR Section 6.2.6. This is essen -
- tially the same test as outlined in Chapter 14 of the FSAR for the preoperational type A test. There are some differences in the test alignments for some penetrations due to post operation system alignment >
needs. The procedure allows for ND to remain in -
operation for core cooling. Additionally, the >
structural integrity test (test pressure at 110 -
115 percent of design pressure) is not performed. ,
As stated in the FSAR and Tech Specs, the test is 36
,- g
performed in accordance with Appendix J of 10CFR50 and the provisions of ANSI N45.4-1972 or the Mass Plot method.
The test is performed by sealing the containment vessel and pumping in air to achieve a pressure of 14.88 psig to 15.0 psig. A four hour stabilization period is then entered. If stable conditions exist after at least four hours, the test is started. A test pressure of 14.88 - 15.0 psig is used to ensure that the containment pressure. remains greater than 14.68 psig (Pa) for the duration of the test.- After completion of the test, a known leak is imposed on the containment vessel to verify the accuracy of the leak rate calculations. Given satisfactory'results, the containment vessel is depressurized and the test concluded.
Overpressure protection of the containment vessel is ensured by the use of a pressure relief valve installed on penetration M346. This valve is set to open at 17.25 psig. The setpoint of this valve will never be reached, however, because the portable compressors will be shutdown at less than or equal to 15.0 psig and VI to containment isolated. Table 3.8.2-9 in the FSAR gives the ultimate internal-pressure for the containment vessel as 72.0 psi.
This test will be well below that value.
Due to the increased air density inside containment when pressurized for the test, the containment fire detectors will go into alarm. A fire watch will be set up using the 52 test RTDs as the means of fire detection. These RTDs are located in lower con-tainment, upper containment and the ice condenser, thus provioing adequate coverage of the entire containment. #
The alignments for this test will result in the requirements for containment closure and containment integrity for fuel movement being satisfied.
Due to the' increased air density inside containment during ILRT, it-is necessary to decrease the speed of the ice condenser (NF) air handling unit (AHV) fans. It will also be necessary to increase the capacity of the thermal overloads for the NF AHU fan
. motors. The pressure increase inside containment will result in a doubling of the' air density. Since fan power is directly proportional to fan rpm cubed times the density, the power required by the fan motor will double. The new sheaves to be installed per this procedure will result in a fan speed which i is 85% of current fan speed. Therefore, with this ,
reduced speed and doubled air density, the required !
17
h 3'
F fan motor power will be 1.25 times the normal value, i p This will be accounted for by increasing the capac-o ity of the thermal overloads for the fan motors to 1.5 times their current capacity. Due to the length
, of time required to install these modifications, ,
installation will start prior to' entering mode 5. I These fans are not-safety related, are not required i for safe shutdown of the unit and have no Tech Specs !
e directly related to them, therefore operability is '
h not a concern. They do, however, maintain the ice !
r condenser at its required temperature. The ice ;
i.
condenser temperature is monitored at least every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as required by Tech Spec surveillance re-quirement 4.6.5.1.a. therefore any problem with ice j condenser temperature would be discovered and l corrective action taken. This test will be logged into the shift supervisor's logbook prior to imple-menting these modification and will remain logged in ;
until the test is complete and the modifications ,
removed. 1 Therefore, the probability and consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. The possibility of an accident or ,
malfunction of equipment important to safety not previously evaluated in the FSAR will not be creat- ,
ed. The margin of safety in the basis to Tech Specs will not be reduced. No.USQ was deemed to exist. ;
PT/2/A/4200/02C-H Containment Closure Verification: PT/2/A/4200/02 was originally used to check all penetrations for Refueling. Containment Integrity. This was found to be a large and cumbersome procedure.
The procedure was later divided up into 6 different procedures, .
PT/2/A/4200/02C thru PT/2/A/4200/02H. These proce-dures were not divided up by penetration location but rather by penetration number and were still difficult to perform. This retype reorganizes the penetration checklist so that the penetration' will :
be grouped by location. Additional changes to this l periodic test are valve location changes, human factor changes of the penetration checklist, delet-ing the S/G PORV Block Valve from'the penetration L checklist for penetrations M-113 and M-423. The S/G PORV is located downstream of the Block Valve and is being used as the isolation valve. Double isolation is not required to meet the acceptance criteria when isolating the penetration. One additional change >
included with this retype is that sections of this ,
procedure have been revised to verify containment is
- capable of being closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at any time while NC System level is less than or equal to 16%.
This change is the result of Generic Letter No.
88-17 and reasonably assures that containment
.]
closure will be achieved prior to the time at which !
p a core uncovery could result from a loss of Decay b Heat Removal. This procedure is not used to manip-ulate any system and is only used as a verification process for penetrations. No VSQ was deemed to exist.
PT/2/A/4200/31 SV Valve Inservice Test - Quarterly: This proce-dure, PT/2/A/4200/31 SV Valve Inservice Test -
Quarterly, is a procedure writer's guide upgrade of an existing procedure. The procedure is used to i satisfy IWV stroke time requirements for valves 2SV1 S/G 2D power operated relief valve (PORV), 2SV7 S/G !
20 PORV, 2SV13 S/G 2B PORV, and 2SV19 S/G 2A PORV.
These valves are designed to: ,
- 1) Prevent lifting of the Code Safeties on each :
steam line during mild transients.
- 2) Assist in reseating the safeties.
- 3) Provide a mens for cooldown when the steam dump "
system is unavailable.
- 4) Provide a. safety grade means of cooldown to ND ;
initiation.
- 5) Close on any main steam isolation signal.
This test is written to be performed in any mode.
The test allows any valve-to be tested individually or all 4 to be tested at once. The valve (s) to be tested has its associated block valve closed,-is opened, then closed by the closure of a switched ,
jumper installed across the SSPS output relay contacts for main steam isolation. From the time. -
this switch is closed until it is removed and the
~
PORV's are reset, NONE of the PORV's will be capable of opening automatically. They will ALL be. capable ,
of opening in manual mode at any time during the-test. Therefore, the safety function in item 4 "
above-is not diminished. The ability to provide :
main steam isolation at any time during the test is- '
not diminished. For any valve (s) tested the closed block valve will provide isolation during the time the PORV is open. If the isolated PORV(s) is needed for cooldown to ND initiation, the block valve can be opened and cooldown initiated.
Therefore, the probability and consequences of an accident or malfunctico of equipment important to safety previously evaluated _in the FSAR will not be .
increased. The possibility of an accident or malfunction of equipment important to safety not previously evaluated in the FSAR will not be 39
created. The margin of safety in the basis to Tech Specs will not be reduced. No USQ was deemed to exist.
PT/2/B/4250/12 Condenser Tube Leak Detection System Operability Test: This procedure is used to periodically test and document the operability of the condenser tube -
leak detection system. It is an administrative procedure to ensure adequate levels of performance during periods of condenser tube leaks. Actual equipment operation is done per OP/0/0/6250/13, Operating Procedure For The Condenser Tube Leak Detection System, a previously approved and evalu-ated procedure. ,
Since all equipment operation is done per a previ-ously approved and evaluated procedure, the proba-bility or consequences of an accident previously evaluated in the FSAR will not be increased, nor will the possibility of an accident not previously evaluated in the FSAR be created. Since no safety related equipment is involved, the probability or consequences of malfunction of equipment important >
to safety previously evaluated in the FSAR will not be increased, nor will the possibility of malfunc-tion of equipment different than previously evalu-ated be created. Since no Technical Specifications are involved, no margins of safety will be reduced.
No USQ was deemed to exist.
PT/2/A/4550/01A Preparation For Refueling: PT/2/A/4550/01A retype
- 3 incorporates changes 3 thru 5 along with additional changes. This retype removes load testing of the Fuel Hoist portion of this PT to PT/2/A/4550/01E (Spent Fuel Building Manipulator Crane Load Test) and the New Fuel and New RCCA Handling Tools to PT/2/A/4550/01F (Preparation for New Fuel Receipt), Numerous editorial changes have been made throughout the procedure. The ma,ior change now directs the operator to the applicable equipment operating procedures to perform the various verifications made. This ensures that as the equipment procedures are updated the need to update this PT is reduced. It also ensures consistency in the operation of the equipment and ensures the Limits and Precautions for the '
applicable equipment are reviewed. Other editorial revisions ensure the operator understands the event to take place following an action prior to the operator performing the action.
The performance of this procedure ensures the intent I of paragraph 9.1.4.4 of the FSAR is met, for the equipment in the Spent Fuel Pool. The use of this procedure does not change the probability or 10
u consequences of any accident or equipment malfunc-tion discussed in the FSAR. Nor does it create the possibility of a different accident or equipment malfunction scenario. There are no Technical Specification Requirements for this equipment, therefore the margin of Safety for the Tech Spec's as defined by the Bases is unaffected. No USQ was deemed to exist.
PT/2/A/4550/01E Spent Fuel Building Manipulator Crane Load Test:
PT/2/A/4550/01E was originally the_ load test portion ofPT/2/A/4550/01A(PreparationforRefueling). It was separated from pT/2/A/4550/01A (Preparation for Refueling) to improve useability of the procedure and to allow simplified retesting of the mast following maintenance to it. Additional changes to the content have been made from the original proce-dure to ensure the operator understands the event to take place following an action prior to the operator performing the action. PT/2/A/4550/01A gave fixed numbers for the Overload, Underload and Lowload setpoints of the crane. These settings are no longer fixed, but are calculated in OP/2/A/6550/06 (Transferring Fuel with the Spent Fuel Manipulator Crane) based on actual weights of Mast and if needed, an irraaiated fuel assembly. The above calculations are based on Westinghouse F5.0 and F5.1 specifications to ensure the safe movement of fuel.
The mechanical overload setpoint has been reduced from 3250 lbs 1 20 lbs to 2540 1 160 lbs to allow a more conservative setting of this back up devices' setting. The allowable span was-increased to allow a more realistic operating band for this mechanical device. See attached letter on requested FSAR update to FSAR Section 9.1.4.
The development and use of this procedure does not change the probebility or consequences of any accident or equipment malfunction discussed in the F3AR. Nor does it create the possibility of a dif ferent accident or equipment failure scenarios.
Load testing of the Spent Fuel Building Manipulator Crane is not required by Tech Spec and, therefore, the Tech Spec Bases are not affected by this procedure, No USQ was deemed to exist.
TT/1/A/9200/51 ATWS Functional Test for NSM CN-10952: This proce-dure change will allow resetting only one CF Pump instead of both, since CF Pump 1B is unavailable at this time. This change will not affect the test-results since both CF pumps must be tripped in order l L
to actuate AMSAC. CF Pump 1B will remain tripped and CF Pump 1A will be reset to clear the AMSAC signal. CF Pump 1A will then be tripped to initiate =
l 41 1
r
[ the AMSAC signal in Section 12.1.1. In Section 12.1.2, CF Pump 1A will be reset to clear the AMSAC signal so that the CF Flowpath Blockage signal can initiate AMSAC.
Since the actual test method is not affected, no j accident scenarios are affected. No new accidents are created. Equipment malfunction possibilities and consequences are actually decreased, since one less piece of equipment is to be operated as a result of this change.
Since none of the systems affected by this change ,
are required in Modes 4, 5, or 6, the margin of !
safety as defined by the bases of Tech Specs. No USQ was deemed to exist.
PT/1/A/4400/06E KD Heat Exchenger 1A Heat Capacity Test: Steps were added to Section 6.0 to allow returning the system to "As Found" conditions if the test can not be completed and to give a limit on RN temperature.
Steps were deleted from the prerequisites in Sec-tions 7.0 and 8.0 that were not needed. Additional steps were added in Sections 12.0 and 13.0 to allow !
measurement of_the KD temperature entering the Lube Oil heat exchanger (for information only) and to N/A ,
these steps if the temperature is not monitored.
Steps were also added to declare the Diesel Genera-tor inoperable if the position of IRN-236 is changed and back to operable after 1RN-236 has been returned to the "As Found" position. All other changes were made to procedure in order to perform with the dLOG ,
data acquisition computer program and verify param-eters are entered correctly into the program, The actual alignment of systems and basic order of procedure was not changed. ;
This procedure measures the shell (KD) and tube side (RN) inlet and outle temperatures as well as tube side flow in order to determine the shell side fouling factor for the KD 1A Heat Exchanger. When RN temperature is low, IRN-236 needs to be throttled back to obtain stable test readings. This position is different from the RN flow balance position, so this makes the Diesel Generator (D/G) inoperable.
Steps were added to notify Operations that D/G 1A is inoperable when the position of IRN-236 is changed and is operable when 1RN-236 is returned to the "As Found" position. Changes were made to use the dLOG computer system to acquire data, perform calcula- .
tions and build the data set for the Design Engi-neering program that calculates the shell side fouling factor. The dLOG computer system is benchmarked and all files used by this test and 4_2
L parameters. input into the program are verified correct by the procedure. Measurement of the KD ,
temperature entering the Lube Oil heat eFehanger was added for information only and does not enter calculations for the fouling factor.
This procedure is performed while the Diesel Gener-ator is operating in the normal lineup. Shell and tube side temperatures are monitored during test to '
ensure they stay within design limits. .D/G 1A is declared inoperable for less than two hours when throttling of 1RN-236 is required and D/G 18 is operable for this period as required by Tech Specs. ;
So the margin of safety as defined in the basis of Tech Spec will not be reduced. This test does not involve any jumpers, sliding. links, unusual align-ments or any other modifications to safety related -
equipment that would create or increase the proba-bility of an accident. Removing test equipment connected to the heat exchanger and replacing of any removed process instrumentation are independently verified within procedure. For these reasons and the ones stated above, this procedure does not create or increase the probability of a malfunction -
of equipment important to safety or increase the consequences of an accident. Therefore, an unreviewed safety question does not exist.
I PT/1/A/4350/16A 'VG System 1A Valves Full Stroke Verification Test:
The first six parts of this change only serve to improve the steps that were already in place. In steps that included more than one action, the steps are broken up. Those that included unnecessary operations signoffs were rewritten.
The seventh part of the change makes our acceptance criteria for an 11-second start more restrictive (10-second). This allows for any human error involved in starting the visicorder.
Steps 8-14 of the change involve the visicorder.
. The procedure was written to use the totalizcr for l timing the start. The totalizers have proven to be l unreliable and have caused many extra diesel starts.
This change adds all necessary information for l setting up the visicorder and timing the start accurately.
This change inproves the readability of the test and the reliability of the data. No systems, struc-l tures, or components will be changed. No safety
- margins will be decreased. No USQ was deemed to exist. ,
i
c , ,
}.
b ;
t ;
I
!I I
, _ PT/1/A/4350/16B VG System IB Valves Full Stroke Verification Test:
- l. The first six parts of this change only serve to improve the steps that were already in place. In steps that included more than one action, the steps are broken up. Those that included unnecessary
[ operations signoffs were rewritten.
The seventh part of the change makes our acceptance criteria for an 11-second start more restrictive (10-second). This allows for any human error j involved in starting the visicorder.
!' Steps 8-14 of the change involve the visicorder.
The procedure was written to use the totalizer for ;
timing the start. The totalizers have proven to be '
unreliable and have caused many extra diesel-starts.
This change adds all necessary information for setting up the visicorder and timing the start accurately.
This change improves the readability of the test and the reliability of the data. No systems, struc-tures, or components will be changed. No safety margins will be decreased. No USQ was deemed to exist. ,
e PT/0/B/4400/01Y RY Fire Protection Flow Periodic Test: This change allows A & B fire pumps to be racked out for this test. A fire pump is unavailable for operation'due :
to high vibration. B fire pump must be racked out 1' in order to avoid having the pump start automatically i? due to the test. The C Fire Pump will be operable and available to the plant for fire protection. If there is a fire, the procedure directs that the i operator be notified to close IRY17 to allow flow to .
be used in the plant. The procedure also directs :
that 8 fire pump be racked in when testing is complete each day. The test method remains un- '
changed. The plant is in a 7 day action statoment per TS 3.7.10.1 while 2RY pumps are racked out. ,
PT/0/A/4400/08, RN Flow Balance for Degraded Mode: The purpose Change #38 of modification CN-50386 is to add 2 simplex strainers in parallel with the normal RN Pump Lube' ,
Injection Strainer. This modification will allow the strainer to be removed from' service.for cleaning .
without making that train of. RN inoperable. This !
retest will verify that the modification had no affect on normal strainer flows,'and will also verify that minimum acceptable flows can be obtair.ed using the new strainers.
This change in no way increases the probability of .
an accident occurring, whether or not it has been
y , 1 r
I.
P
]
evaluated in the FSAR. This train of RN will not be
- supplying cooling water to inservice components, so {
L no interaction with the reactor or supporting i
- systems is present. The consequences of an accident 1 i previously evaluated in the FSAR are not increased, )
y since the opposite train of RN will be in service '
and supplying cooling water to required components. ,
S This change does not increase the probability of nor consequences of a malfunction of safety related
. equipment. Before valving out e,ny strainer, the strainer being tested will be placed in service. i This will ensure that adequate. bearing injection and motor cooling water flows are maintained at all +
times. No new equipeat failure malfunctions are ;
created as a result of this change. .
The train of RN being tested will be inoperable at -
the time of the test. Technical Specification 3/4.7.4 allows two RN pumps to be out of service for ,
up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in Modes 1, 2, 3, and
- 4. This change will not compromise the operability of the opposite train pumps. Therefore,-the margin of safety as defined in the bases of Technical.
Specifications is not reduced. No USQ was deemed to exist.
OP/0/A/6400/060 Nuclear Service Water System, Change #83: This change incorporates NSM CN-50386. See the 50.59 evaluation for this NSM.
IP/0/A/3010/09E, Main Feedwater (CF) System Miscellaneous Maintenance Changes 0 to 1 Inc. of Feedwater Isolation Valve Actuators: This-procedure is a miscellaneous maintenance procedure for taaintenance such as general inspection and cleaning, and to satisfy EQRI requirements such as hydraulic fluid sampling and replacement, and priming the pump after fluid replacement. The only section in this procedure requiring the USQ evalua-tion is Section 10.8, on action to take during excessive hydraulic / pump motor cycling. This section of the procedure requires power to be removed from the motor circuit long enough to allow hydraulic pressure to fall further than the hydrau-lic pressure switch setting, but not long enough to allow the valve to drift closed. As power is restored, the prolonged pump operation provides flow for a longer period of time to possibly fully seat the component causing the internal hydraulic leak.
The capability of the valve to close in the required 5 seconds is not affected. Performing this section of the procedure is bound by the LOSS OF NORMAL FEEDWATER section of the FSAR Accident Analysis (15.2.7).
15
l The consequences or probability of an accident '
evaluated in the FSAR is~ not increased by this procedure since it does not place the plant outside the design base envelope. The possibility of creating an accident not previously evaluated is not i increased because this procedure does not cause upsets or transients not bound by the FSAR. Also, ,
the probability of a malfunction of equipment previously evaluated in the FSAR is not increased or created. The consequences of a malfunction of equipment important to safety evaluated in the FSAR is not affected. The margin of safety defined in the basis of the Technical Specifications for Catawba is not reduced. . Based upon this evaluation, the FSAR is rot affected and an unreviewed safety question does not exist as a result of this proce-dure.
PT/2/A/4150/26 Adm %!ctrative Controis-for Periodic Testing of Unit >
2 Check Valves: This procedure revision added the test dates for the first and second refueling outage on Unit 2. These changes will not affect the test method or frequency. No unreviewed safety questions ;
are involved. This procedure is used only for -
Administrative Controls of the check valve testing program. No USQ was deemed to exist.
pT/1/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test: ,
Change #126 This procedure modifies the Auxiliary Safeguards ,
procedure so that.the hydrogen skimmer fans may be run without opening the inlet valves to the fans.
Neither the probability nor the consequences of an accident will be created by this change. The change made by this change ensures that in the event of a ,
LOCA during the test, the ice condenser would not be bypassed by air / steam going through the hydrogen ,e skimmer fans, The possibility of an accident r different than any already evaluated in the FSAR ,
will not be created. One train of VX is assumed to be operable in the FSAR and the train opposite the ore being tested will remain unaffected by this ,
test. Neither the probs.biltty nor the consequences :
of a malfunction of equipment will be increased by this change. The hydrogen skimmer fan is run with '
the inlet valve closed every quarter for testing as ,
required by Tech. Spec. 4.6.5.6.lc. The mergin of ,
safety as defined in the bases to Tech. Specs, will not be reduced, pT/2/A/4450/050, Containment Air Return Fan 2B and Hydrogen Skimmer
- Reissue, Changes Fan 2B Performance Test: The use of " TEST" selector -
O to 11 Incorporated switches was deleted from this procedure and a jumper was placed (when applicable) to serve a function of the deleted switch. Deleting the above
, L6
k switches (in some sections) operates the system closer tc, or in.the designed alignment (see below).
When the same terminals at 2SSPSB are used in more L than one section of this procedure, a jumper with a switch is installed and the position of the switch :
is changed to initiate the appropriate action. '
Using the jumper with a switch reduces the number of times that jumpers are placed in 2SSPSB so the possibility of a personnel error is reduced. ,
I Section 12.1 of this procedure verifies the auto-start of the Air Return Fan (ARF-2B) after an Sp test signal, CPCC permissives to start and stop the !
fan, the 15 minute run time, and the fan running -
o speed and current with the discharge damper (ARF--
D-4)~ closed. The order of steps were changed to verify the CPCC permissives for the fan. Steps were added to Section 12.1 to remove power from damper a ARF-D-4, af ter being verified closed, before' the f an (ARF-28) is started and return the power after the fan is shut down. This change reduces the possi- l bility of opening ARF-D-4 and inadvertently opening the ice condenser doors during operation of the Air !
Return Fan. Also, in the event of a LOCA during the performance of this Section (12.1) the damper would not open (with the ARF-20 operating) and violate the ,
9 minute time delay for operation of the ARF-2B as assumed in the Design Basis Accident (DBA). The i
jumper idded in 2EATC2 serves the function (opening the bypeiss dampers when ARF-2B starts) of the deleted " TEST" selector switch (VX49). Without the .
use of !.he " TEST" selector switch and with the new ,
order of steps to verify the CPCC permissives, the ARF-2B is run after the Sp test signal is removed
[ and then the CPCC permissive is verified to stop the fan. The above changes operate the ARF-2B closer to a the designed accident alignment, ,
Section 12.2 of this procedure verifies the auto-open of the Air Return Fan discharge damper (ARF-0-4), on an Sp test signal, after the required time y delay and the CPCC permissives to prevent / enable t.he t opening of ARF-D-4. The iumper added in 2EATC2 was added to verify the CPCC permissives for the damper.
An additional jumper was not added to serve a function (blocking 2VX2B from opening) of the deleted " TEST" selector switch (VX56). Instead, a j CAUTION statement was added to ensure that the Sp test signal is removed before 2VX2B is given the signal to open. Changes were also made.to close 2 ARF-D-4 immediately af ter it opens in order to reduce the possibility of Ice Condenser bypass leakage in the event of a LOCA during the test. The ARF-2B and HSF-20 are blocked from starting by placing their selector switches in the "0FF" position.
AZ f
p- g N
Section 12.3 of this procedure verifies the auto-open of the Hydrogen Skimmer Fan (HSF-28) suction valve (2VX2B), on an Sp test signal, followed by the '
start of HSF-2B after the required time delay.
L Acceptance criteria 11.6 was modified to require L verification of the start time delay for the HSF-28 L which is a conservative change because the start time delay is also verified in acceptance criteria 11.7. An additional jumper is placed in 2SSPSB to initiate an Sp test signal to auto-open 2VX2B and start HSF-28. The above mentioned jumper serves a function (opening of 2VX28) of the deleted " TEST" selector switch (VX52), and also opens 2VX2B using the proper " safety" circuitry. This change operates the HSF-2B in the designed accident alignment. A CAUTION statement was added to ensure that HSF-2B is
[ not operated more than 5 minutes with 2VX2B open.
Changes were also made to stop HSF-28 immediately after it starts and to close 2VX2B after it opens.
These changes reduce the possibility of Ice Con-denser bypass leakage in the event of a LOCA during the test.
The performance of Sections 12.2 and 12.3 of this procedure put the VX system in a configuration which deviates from the assumed initial conditions for a Designed Basis Accident (DBA) because the deck leakage area is increased by opening ARF-D-4 and air is moved from Lower to Upper containment by running HSF-20 with 2VX2B open. In both of these alignments a condition exists for Ice Condenser bypass leakage during a LOCA. The open damper (ARF-D-4) is ad-dressed in PIR 2-087-0061, where the A train damper (ARF-0-2) was found open, and the conclusion of the PIR was that the effect of the additional steam l bypass area on initial compression peak pressure was '
small. Both sections are required by Tech Specs
, 4.6.5.6 to be performed. Tech Specs 3.6.5.5 allows an equipment batch to be open for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> which also 3 deviates (in the same manner as above) from the j initial conditions for a DBA. Changes were made in each section to minimize the time that the system is in an alignment as n.entioned atove, Since the changed test duretion for both Sections 12.2 and 12.3 is on the crder of a few minutes not hours, the margin of safety as defined in Tech Spec will not be reduced.
Section 12.4 of this procedure verifies the auto-start of HSF-20 on an Sp test signal (after the required time delay), the 15 minute run time and the fan running speed and currents with 2VX2B closed.
Additional jumpers are placed to initiate the Sp test signal that starts HSF-2B and to allow opera-tion of HSF-2B with 2VX2B closed. In the event of a ;
41 !
l 3
,y LOCA during the performance of this Section D 2.4),
[c
~
- the HSF-2B would function as designed becasi.e flow ;
would not commence until 2VX2B opens after the Lo, designed 9 minute time delay. The modifications to Section 12.4 were made primarily to include verifi-C a cation of acceptance criteria 11.7, but since no
, " TEST" selector switches were used and an Sp test signal is initiated and then removed after the fan e
- starts, the system is operated closer to the design '
[ accident alignment. :
- ' The Containment Air Return Fan 2B (CARF-28) and Hydrogen Skimmer Fan 2B (HSF-2B) are declared ,
inoperable during the performance of this test. The CARF-2A and HSF-2A will remain operable for the duration of the test as required by Tech. Spec. ,
None of the above changes will prolong the time of !
inoperability or place the systems in any unusual alignments that would create or increase the proba-- '
bility of an accident. The margin of safety as defined in the bases of Tech Spec will not be reduced. Installation / removal of jumpers, open- ~1 ing/ closing of breakers and returning selector switches to the "As Found" position are indepen- -
dently verified within Section 12.0 or 13.0 of the procedure. For these reasons and the ones stated above, these procedure changes do not create or ,
increase the probability of a malfunction of equip-ment important to safety or increase the conse- ..
quences of an accident. Therefore, an unreviewed 1 safety qu?stion does not exist. l pT/1/A/4450/058 Containment Air Return Fan IB and Hydrogen Skimmer Reissue, Changes Fan IB Performance Test: The use of "1EST" selector ;
0 to 22 Incorporated switches was deleted _from this procedure and a :
jumper was placed (when applicable) to serve a function of the deleted switch. Deleting the above switches (in some sections) operates the system closer to or in the designed alignment (see below).
When the same terminals at ISSPSB are used in more than one section of this procedure, a jumper with a switch is installed and the position of the switch is changed to initiate the appropriato action.
Using the jumper with a switch reduces the number of q times that jumpers are placed in ISSPSB so the possibility of a personnel error is reduced.
H Section 12.1 of this procedure verifies the auto-start of the Air Return Fan (ARF-1B) after an i Sp test signal, CPCC permissives to start and stop the fan, the 15 minute run time, and the fan running speed and current with the discharge damper (ARF-D-4) closed. Steps were added to Section 12.1 to remove power from damper ARF-D-4, after being M l l
verified closed, before the f an (ARF-10) is started and return the power after the tan is shut down.
This change reduces the possibility of opening ARF-D-4 and inadvertently opening the ice condenser doors during operation of the Air Return Fan. Also, in the event of a LOCA during the performance of-this Section (12.1) the damper would not open (with the ARF-1B operating) and violate the 9 minute time delay for operation of the ARF-1B as assumed in the i Design Basis Accident (DBA). The jumper added in i 1EATC2 serves the function (opening the bypass i dampers when ARF-)B starts) of the deleted " TEST" t selector swtich (VX49). Without the use of the
" TEST" selector switch mentioned above, the ARF-1B <
is run after the test Sp signal is removed and then the CPCC permissive is verified to stop the ARF-18. l The above changes operate the ARF-1B closer to the designed accident alignment.
Section 12.2 of this procedure verifies the auto- '
open of the Air Return Fan discharge damper (ARF-D-4), on an $p test _ signal, after the required time delay and the CPCC permissives to prevent / enable the t opening of ARF-D-4. An additional jumper was not #
added to serve a function (blocking IVX?B from )
opening) of the deleted " TEST" selector switch (VX56). Instead, a CAUTION statement was added to ensure that the Sp test signal is removed before }
IVX2B is given the signal to open. Changes were also made to close 1 ARF-D-4 immediately after it opens in order-to reduce the poss_ibility of Ice Condenser bypass leakage in the event of a LOCA during the test. The ARF-1B and HSF-1B are blocked from starting by placing their selector switches in-the "0FF" position, i Section 12.3 of this procedure verifies the auto-open of the Hydrogen Skimmer Fan (HSF-1B) suction valve (1VX28), on an Sp test signal, followe permissive met. The actual-damper position is closed, however. In accordance with Tech Specs, fan.
motor current and voltage are measured and the fan is run for 15 minutes. If a LOCA were to occur during this test, the "A" train VX system will be able to perform its function. Although the fan will already be running, flow will not commence until 2VX1A.is opened. "A" train of VX is technically inoperable since the HSF-2A run permissive from 2VX1A will be met without the damper actually being open. As stated above for other procedure sections, the Minimum Engineered Safety Features Performance 1s maintained, since "B" train of VX will be opera- >
l ble.
Based on the above evaluation, no Unreviewed Safety .
Question exists. !
PT/2/A/4600/08A Verification of QPTR Using Incore Detectors:
Changes included in this retype:
New Test Method Obtain reference incore from most recent FCFM from ,
PT/2/A/4600/05A, Core Power Distribution, y Take a surveillance map (either a FCFM or an '!
eight-symmetrical locations map). Using SNACORE generate the surveillance incore tilts.
Calculate incore QPTR = (Surveillance incore tilt minus the reference incore tilt plus 1.0) for each quadrant. This is the deviation in incore tilt which is equivalent to QPTR. e At the same time the surveillance map was being made, excore QPTR data was being gathered. If Incore QPTR and Excore QPTR $ 1.02, then consistency -
has been shown betwee_n the normalized symmetric power distribution and the excore-QPTR, The majority of the changes included with this retype are related to the new test. method or to. -
upgrade the procedure per the Procedure Writers ;
Guide.
Title Change to " Verification of QPTR Using Incore Detectors" to be more descriptive of what the procedure is used for.
58
=- -
7 n
1.0 - Rewritten to better describe- the purpose of the procedure.
2.0 - References expanded to show all Tech-Specs, FSAR, and procedures relevant to this procedure.
6.0 - Added new Limit and Precaution concerning contacting Health Physics before withdrawing ENA Detectors from storage.
- Deleted reference to PT/2/A/4600/05A, Incore and NI System Correlation Check since new method does not use this procedure.
t 8.0 - Deleted requirement for boron samples since not used, i
- Added NC and PZR pressures to stability requirements.
- Added WARNING concerning notifying Health .
Physics before withdrawing ENA detectors from storage.
- Added step to record when Power Range -
Detector was declared inoperable.
- Added step to determine if the surveillance map can be an eight-symmetrical locations map.
9.0 - Rewritten Test Method to better describe what the procedure does.
10.0 - Rewritten Data Requirements to accurately describe the data used in the new test ,
, method.
11.0 - Rewritten Acceptance Criteria to be' consis-tent with-new test method.
12.0 - Rewritten step to perform surveillance of Tech Spec 4.2.4.2 using new test method,- and
. Procedure Writer's Guide philosophy.
13.0 - Deleted unnecessary. enclosures that are not used with new test method.
The purpose of this procedure-is to perform the surveillance requirement ~of Tech Spec 4.2.4.2 when-above 75% Rated Thermal Power with one Inoperable power Range Detector by use of the Incore Instru-mentation System (ENA).
5.9 ,
7 This test is bound by Condition I event: Operation with permissible deviation, testing as allowed by Tech Spec in scope of ~a normal surveillance re-quirement. So-the probability and-consequence of an accident already evaluated.in the FSAR will not increase, and there would be no new accident sce-nario created by the use of this procedure.
This procedure uses'the ENA System as it was in-tended and designed for and does not place any safety equipment in an off-normal condition.
Therefore, the probability and consequence of an equipment malfunction important to safety as already evaluated in-the FSAR will not increase, and there would be no new possibility of safety equipment malfunction not already evaluated in the FSAR.
The procedure performs the Tech Spec surveillance within the scope of the bases to Tech Specs.
Therefore, the margin of safety as defined in the bases to Tech Specs will not be reduced.
PT/2/A/4400/06E KD Heat Exchanger 2A Heat Capacity Test: The KD Heat Capacity Test determines the tube side fouling factor of the KD Heat Exchanger. This is done by measuring the inlet and outlet temperatures on both the KD and RN side of the heat exchanger, and measuring the RN flow. Data is taken using a dLOG data acquisition computer system. All programs and calculations in the dLOG computer system are benchmarked. Results are calculated using a Design Engineering computer program that has been approved for analysis of the heat exchanger results.
During the KD Heat Capacity Test. the Diesel = Genera-tor is operated in the normal lineup. RN flow may be throttled to allow temperature readings to be more stable. If RN flow-is throttled, this will have no significant~ effect on the plant because the diesel generator will be declared inoperable until the valve is returned to the as found position.
There will be no danger of damaging the diesel during the period of time when the RN flow is throttled, because KD temperature is monitored and verified to be less than 190 F throughout the test.
pT/2/A/4400/06F The KD Heat Capacity Test determines the tube side fouling factor of the KD Heat Exchanger. This is done by measuring the inlet and outlet temperatures on both the KD and RN side of the heat exchanger, and measuring the RN flow. Data is taken using a dLOG data acquisition computer system. All programs and calculations in the dLOG computer system are benchmarked. Results are calculated using a Design Engineering computer program that has been approved for analysis of the heat exchanger results.
f_0
p > .
[ ,
E During.the KD Heat Capacity Test, the Diesel Genera- 1 tor is. operated in'the normal lineup. RN flow may I be throttled to allow temperature readings to be-more stable. If RN flow is throttled, this will have no significant effect on,the plant because the R diesel generator will be declared inoperable until 4
_the valve is returned to the as found position.
There will be no danger of damaging the diesel
_L during the period of time when the RN-flow is throttled, because KD temperature is monitored and -!
verified to be less than 190?F throughout the test.
PT/1/A/4600/08A Verification of QPTR Vsing Incore Detectors:
Changes included in this retype:
New Test Method Obtain reference incore from most recent FCFM from PT/1/A/4600/05A, Core Power Distribution.
Take a surveillance map (either a FCFM or an eight-symmetrical locations map). Using.SNACORE
_ generate-the survei_11ance incore tilts.
Calculate incore QPTR = (Surveillance incore tilt minus the reference incore tilt plus 1.0) for each quadrant. This is the deviation in incore tilt which is equivalent.to QPTR.
At the same time the surveillance map was being.
made, excore QPTR_ data was being gathered. If .
Incore QPTR and Excore QPTR 5 1.02, then consistency
.has been shown between the normalized symmetric power distribution and the excore QPTR.
The majority of the changes . included with this retype'are related to the new test method or to upgrade the procedure per the Procedure-Writers Guide.
Title Change to " Verification of QPTR Using Incore ;
Detectors" to be more descriptive of what the procedure is used for. i 1.0 - Rewritten to better describe the purpose of the procedure.
2.0 - References expanded to show all Tech Specs, FSAR, and procedures relevant to this procedure.
6.0 - Added new Limit and Precaution concerning contacting Health Physics before withdrawing ENA Detectors from storage.
61
E n -
- Deleted reference to PT/1/A/4600/05A, Incore <
and NI System Correlation Check since.new method does not use this procedure. '
8.0 - Deleted requirement for boron samples since not used. !
- Added NC and PZR pressures to stability ;
requirements.
- Added WARNING concerning notifying Health Physics before withdrawing ENA detectors from storage. !
- Added step to record when Power Range n" Detector was declared inoperable.
- Added step to determine if the surveillance map _can be an.eight-symmri.rical locations map.
9.0 - Rewritten Test Method to better describe what.
the procedure does.
10.0 - Rewritten Data Requirements to accurately 1 describe the data used in the new test method. i 11.0 - Rewritten Acceptance Criteria to be consis-
' tent with new test method.
12.0 - Rewritten step to perform surveillance'of Tech Spec 4.2.4.2 using new test method, and Procedure Writer's Guido philosophy.
- 13.0 - Deleted unnecessary enclosures that are not i
used with new test method.
The purpose of this procedure is to perform the
-surveillance requirement of Tech Spec 4.2.4.2 when- 1 i- above 75% Rated Thermal Power with one Inoperable Power Range Detector by use of the Incore Instru-mentation System (ENA).
This test is bound by Condition I event: Operation with permissible deviation, testing as allowed by Tech Spec in scope of a normal surveillance re-quirement. So the probability and consequence of an '
accident already evaluated in the FSAR will not increase, and there would be no new accident sce-nario created by the use of this procedure.
4
Nw
(
This procedure uses the ENA System as it was in-K tended and' designed for and does not place any safety equipment in an off-normal condition..
Therefore, the probability and consequence of an equipment. malfunction important to safety as already evaluated in the FSAR will not increase,-and there would be no new possibility of safety equipment- -
malfunction not already evaluated in the.FSAR. j The procedure performs the Tech. Spec surveillance within the scope of the bases to Tech Specs.
Therefore, the margin of safety as defined in the bases to Tech Specs will nut be reduced.
Mp/0/A/7150/44 Safety Injection Pump Corrective Faintenance: This ;
safety evaluation is for the revision of-MP/0/A/7150/44 change 11, prepared on 05-15-89.
This procedure is being reviced to incorporate- '
changes to the instruction manual.and correct errors identified in the procedure. These revisions and >
changes will increase the reliability of the work controlled by this procedure. The changes to this '
procedure are identified on the procedure major change form.
.i Tech Specs 3.5.2 are affected by this procedure. L Operations has the respon.ibility and the procedures ,
for. compliance with this Tech. Spec. Maintenance !
will be performed on this pump when Tech. Specs, allow, per Operation's procedures. These revisions will clarify and assure that maintenance activities will return the Safety' Injection Pump to as-designed conditions.
1 A full USQ evaluation is required because this i,
procedure is being changed significantly and it is ,
described in the FSAR.' FSAR Section 13.5.2.2.1 :r (Maintenance Procedures) states that maintenance procedures are required and since this procedure change, as described above, changes this procedure in a significant manner, a USQ evaluation must be ,
performed.
The corrections made by this procedure change have been reviewed against approved vendor manuals, design documents, and station proceoures to ensure
< that the corrective maintenance controlled by this procedure will return the pump to as-built /as-designed
- condition. These actions will ensure the pump's ccmpliance with FSAR accident analysis. Since the pump will be returned to as-designed conditions, the i possibility, consequences, or probability of a malfunction will be reduced. The possibility, probability or consequences of a previously unreviewed h g3
i safety question are not created by this change-because the pump will be returned to as-designed ')
conditions. By the same reasoning, the margin of-s safety as defined in the Bases of the Tech Specs will not-be reduced. Therefore, no USQ exists.
PT/1/A/4200/33A VI Valve Inservice' Test - Quarterly: This proce- i dure, PT/1/A/4200/33A, VI Valve Inservice. Test -
Quarterly, is a Procedure Writer's Guide upgrade of '
an existing procedure. -This procedure is used.to satisfy Technical Specification 4.0.5 stroke time requirement in the time specified by Tech Spec Table 3.6-2a. The valve stroke timed is IVI-312A which is ,
a Unit 1 Instrument Air Supply to the Containment Purge Valves Containment Isolation. This valve is ;
required to be stroked from open to closed.
This valve receives a containment isolation T signal (Containment High Pressure) to close. 1VI-312A is timed closed by using the control room pushbuttons.
The Instrument Air'(VI) system supplies clean, oil free, dried air to all air-operated instrumentation-and valves. During modes 1-4, the VP valves are
- closed with power removed. -Isolating VI to these valves would only fail them closed. Below mode 4, the VP system is not needed. FSAR Section 9.3.1.3-states ' Failure of the: compressed air systems will not render any safety system equipment or'its function inoperable. A loss of instrument air :i during an accident or station blackout would cause all pneumatically operated valves in the station which are essential for safe operation to fail-in the safe position..
The probability and consequences of an accident or malfunction of equipment -important to safety.previ-' .
ously evaluated in the FSAR will not be increased, d-The possibility of an accident or malfunction'of equipment important to safety not previously -evalu-ated in the FSAR will not be created. The margin of- <
safety in the bases to Tech Specs will not be I reduced.
PT/1/A/4200/24 NF Valve Inservice Test - Quarterly: This proce-dure, PT/1/A/4200/24 NF Valve Inservice Test,.is a Procedure Writer's Guide upgrade of'an existing procedure. This procedure is used to satisfy .
Technical Specification 4.0.5. stroke time require-ment in the time specified by Tech Spec Table 3.6-2a. The valves stroke timed are INF-228A, INF-233B, and INF-234A which are Unit 1 Air _ Handling Units Glycol Supply or Return Containment Isolation Valves. These valves are required to be stroked from open to closed. Also, INF-228A and 1NF-234A are verified to fall closed.
6A
These valves receive.a containment isolation T signal (Containment High Pressure) to close.
INF-233B is timed closed by using the Control Room pushbuttons. INF-228A and 1NF-234A are closed by opening a sliding link. Each link affects one valve only and the valve. fails closed to its' safe posi-tion. Prior to performing any of the valve strokes, the glycol pumps and chillers are shutdown by Operations. After performing the valve strokes, Operations returns the glycol system to service.-
According to FSAR 6.7.6.1, "During a postulated loss-of-coolant accident the Refrigeration' System is not required to provide-any heet removal function" and FSAR 6.2.1.1.2, "A particular advantage of the ice condenser is its passive actuation not requiring an actuation. system signal". The removal from
. service of the NF chillers and pumps does not affect operability of the ice condenser.
The probability and consequences of an accident or malfunction of equipment important to safety previ-ously evaluated in the FSAR will not be increased..
The possibility of an accident or malfunction of equipment important to safety not previously evalu-ated in the FSAR will not be created. The margin of safety in the bases to. Tech Specs will.not.be reduced, pT/2/A/46p0/10 Incore Detector Thimble Eddy Current Testing: .This procedure directs the eddy current testing of the Incore Detector system flux thimbles. It has no effect on any plant structures, systems or compo-nents mentioned in the FSAR with the exception of the incore thimbles themselves which are part of the Reactor Coolant System pressure boundary. Passage of a test probe the length ~of each incore thimble will have no more effect than.the. passage of the Incore Detectors during normal system operation in Modes 1 or 2, (since this probe is approximately the same diameter as an Incore Detector). This testing-will be conducted in Mode 5 with the NC System depressurized to less than 5 psig. The overall effect on these components is therefore negligible.
The purpose of these measurements is to assess the magnitude of vibration induced thimble wall thinning which has been identified as an industry wide concern in Westinghouse reactors by NRC Bulletin No.
88-09. Based upon the evaluation of the eddy current results, thimbles with excessive through-wall wear shall either be isolated at the seal table or repositioned to move the region of thimble wear out of the area where wear occurs (the area between the lower core plate and the fuel assembly incore 61
e instrument guide tube). This testing program is :r being undertaken to anticipate (and take action to prevent) potential thimble failures, which are breaches in the Reactor Coolant System pressure boundary.. This testing has been performed on Unit
- 1. This procedure differs from the Unit 1 procedure
~, only in that Thimble Cleaning is not required before testim s long as thimbles have been shown to be
( unobsm :;ted (for example, through flux mapping).- ,
- Incore thimble eddy current testing will not in-crease either the probability or the consequences of any accident previously evaluated in the FSAR. '
Section 15.6.5, Loss-of-Coolant Accidents, has been evaluated from the standpoint of a small break (<
1.0 ft") LOCA induced by the eddy current test. probe causing a failure of a thimble with excessive through-wall wear during testing. This unlikely scenario is bounded by the small break LOCA analysis since it would occur in Mode 5.with the NC System depressurized to less than 5 psig (Section 15.6.5 analysis has been performed assuming Hot Full Power conditions). In this configuration, reactor coolant leakage out of a ruptured thimble through the seal r table would not be of sufficient magnitude to be l unisolatable by placement of a cap on the seal table fitting. Prior to unit restart, the isolation of any ruptured thimble _shall be. qualified as an
- acceptable NCS Pressure boundary component. There is-no possibility of thimble ejection occurring as a result of high pressure fitting failure at the seal table with the NCS at such a low pressure. This assessment was made on'the basis of discussions with-cognizant Westinghouse personnel (Randall Holmes and.
Larry Potochnik) and the inherent physical geometry of the incore F.iimbles with respect to interfacing .
Reactor vessel components. The snug-accommodations of the incore thimbles within the. lower Internals
. guide tubes and Fuel Assembly guide tubes disperses the force upon the thimbles posed by a differential-pressure of 5 psig to such an extent that ejection-is not possible. Any such fitting failure can be corrected at the seal. table with negligible imped-ance by NCS-leakage associated with it.
Neither the probability nor the consequences of a malfunction of equipment important to safety, as previously analyzed in the FSAR, will be increased by Incore thimble eddy current testing. The only equipment involved are the-Incore Detector thimbles themselves. A worst case sceiario of a thimble rupture resulting from the aggravation of an already weakened thimble by passage of the test probe (or the failure of eddy current testing to identify a 6p A_.. .__z _ _ . _ . _ . . _ __ __.
/
significantly degraded' thimble) following unit:
restart is bounded by FSAR analyses.- The isolation of leaking thimbles is addressed specifically.in Section 7.7.1.9, Incore Instrumentation, which states that such action-can be performed in Mode 1 or, at worst, with the unit shutdown in Mode 5.
The. possibility of an accident or equipment mal-function not previously analyzed in-the FSAR will not be. increased by this test program. The purpose of this testing is to minimize the probability of.
previously analyzed situations, which are all inclusive with respect to the Incore Detector System thimbles,.from occurring.
The margin of safety as defined in the applicable.
Tech Specs will not be degraded by this' testing.
The bases of Tech Specs 3/4.2.1, 3/4.2.2, 3/4.2.3 and 3/4.2.4 are all critically affected by avail- 3 ability of the Incore Detector System. Reposition-ing of degraded Incore thimbles will be performed as !
.part of this testing as necessary to ensure system operability (> 75% of instrumented core locations <
available).
- EP/2/A/5500/19, Loss of Residual Heat Removal System: Evaluation of the revisions to AP/2/A/5500/19, Loss of Residual
~
Retype #8 Heat Removal System, based on the recommendations of Generic Letter 88-17 " Loss of Decay Heat Removal".
In response to the recommendations of Generic: Letter 88-17 " Loss of Decay Heat Removal," AP/2/A/5500/19, Loss of hesidual Heat Removal System, has been revised. Through-close cooperation ~of the NRC Project Manager for this concern and'the Westing- :
house Owners Group (WOG), interim guidance, WOG U 88-156 has been.provided to the individual _ owners'on -
preventing, identifying and recovering from a loss of decay heat removal capability. The majority of the changes made are' based on this interim guidance.
AP/2/A/5500/19, Loss of Residual Heat Removal ~
System, formally consisted of two cases which-functioned to guide the operator through the loss of an ND train or a leak in the ND System. The current revision further subdivides the loss of an ND train based on the status of the NC system pressure boundary. If the ND pressure boundary is intact.and capable of being pressurized, then Case 1 applies.
If the NC pressure boundary is open to Containment atmosphere and is not capable of being pressurized, then Case III applies.
bl
Case II which' covers a leak in the ND System has
, been revised to utilize the alternate means of residual heat removal ~of either Case I or III based on plant conditions. The means of determining the 1eaking train-has been enhanced by this revision.
Case I Loss of ND Train with NC Pressure Boundary 1 Intact. ,
Revisions based on Westinghouse inter'im guidance are well founded as documented in the FSAR and Westing-
. house Emergency Response Guidelines.
The division of' cases based on NC System pressure e boundary status is a direct result of the require-ment to maintain the core covered and preferably-subcooled (stable) under all conditions when resid-ual heat removal capability is lost. ~ The diverse guidance for maintaining the core covered and stable +
under various NC System configurations dictates the .
division of the recovery procedure to facilitate operator use.
Use of the secondary heat sink (CA System and available' steam release capabilities)'as a substi-tute for the normal ND cooling is addressed in FSAR Section-5.4.7.2.5. The secondary heat sink with adequate-makeup to offset normal leakage stabilizes the NC System in the temperature region of 350 Deg F to 212 Deg F-in a subcooled condition for an ex-tended period until the ND System can be restored to
normal operation.
Feed and bleed of the NC System is addressed in Westinghouse Emergency Response Guidelines (FR-H.1) and'specifically in EP/2/A/5000/2C1 Loss of Secon- .
dary Heat Sink. The process for. removing core- L; residual heat at less than 200 Deg F .is'similar to ,
that at high temperatures except no credit is taken for the cooling effect of vaporization. Feed capacity is variable and available for use based on administratively protected makeup systems. If normal makeup capacity is not sufficient to control-NC temperature, then a PZR PORV is opened to relieve NC pressure and high volume pumped or gravity flow makeup is aligned.
Makeup capacity to keep the core covered is based on plant experience for gravity flow to the core during ,
previous refueling outages. With the NC System depressurized, static head of the FWST provides sufficient driving force to maintain flow to the Core.
I j_8
y i 1
Use of PZR PORVs as a bleed path is addressed it, the bases of Technical Specification 3.4.9.3 and SSER Section 5.4.4.1. Capability of relieving NC pres- '
sure during high pressure pump injection.and the -
conditions of Containment after PRT rupture disk failure are discussed.
Containment closure is only required if feed and .
bleed is initiated since the NC System is otherwise intact providing one integral barrier against the o release of fission products to the environment.
Under normal cooldown conditions, Containment-is not opened until the NC System is less than 200 Deg F.
Case II Leak 'n ND Procedure changes in Case II functionally aligned the procedure for use with Case I or Case III based ;
on plant conditions. Case II remains an entry point into the Ap when decreasing NC inventory is indi-cated. Immediate action to stop the operating ND pump (s) requires branching to the ND restoration procedures.
Subsequent steps in Case II determine the train affected by the. leak, isolate the affected train and.
determine what component in the affected train is leaking. The end of Case II is the return to Case I or III to restore normal ND operation or maintain the core stable.
The sequence of steps to determine the leaking train was changed to separate the ND trains from each '
other prior to realigning the-suction ~ valves to the NC System. This was done.to better determine which train has the leak. Former procedural guidance-was inadequate. This change is considered an enhance-ment.
Case III Loss of ND Train with NCS Open to Contain-ment Case III consists of two major actions based on Westinghouse interim guidance. The first series of actions are taken to protect the malfunctioning pump ,
and restore normal ND core cooling with the standby ND pump. This single pump operation at reduced NC system inventory is administrative 1y controlled to-minimize the impact of a loss of one ND train. The second series of actions are taken if normal ND cooling with one ND train cannot be restored.
59
e ,
The second series of actions prepares Containment.
and the NC System for " fill'and spill core cooling".
Containment closure and evacuation are. initiated followed by the alignment of a variety of assured makeup f' - aths. The flowpaths are administra- -i tively pru weted to ensure their availability for operator use.
Fill and spill,is very similar to feed and bleed as addressed in Case I, however, the release' path from the NC System is at a much lower elevation. . Typi-cally with the NC. System open, coolant will exit the system at either the S/G manways (11%) or the NC. :
pump seals (16%) to the -lower Containment floor. If the Rx vessel head is removed and S/G nozzle dams are= installed, then the Rx vessel flange.is over- '
flowed to the refueling cavity. Continuous-makeup flow is provided to maintain the core stable, ,
Stability is defined as core exit temperature less than 200 Deg F, when core exit thermocouples and connected, or NC level greater than or equal to -'
overflow elevation with continuous makeup flow (core-covered),
i The remainder of Case III ensures Tech Spec and i reporting requirements are satisfied and ensure: the- f ND System is restored to normal. operation or the ,
core is maintained in a-stable condition by alter-nate means.
When restoring an ND Train to normal operation under the above listed scenarios, certain flow and tem- '
perature limits will apply to ensure that flashing does not occur in the highpoint of the ND pump suction line. Guidance was added to the ND pump restoration section of Case I and Case III to ensure
.that adequate' net positive suction head is available ;
for.the ND pump when it is~ returned to service.
These flow and temperature limits were provided by Design Engineering in a letter from R. C. Bucy to L.
A. Reed dated March 14, 1989.
Based on the revisions made to the procedure an'd the information sited, the probability of an accident previously evaluated in the FSAR is not increased and the consequences of an accident previously. .
evaluated in the FSAR are not increased. A LOCA is t the closest accident resembling a loss of ND in this drained down configuration. Considering the low energy of the Rx coolant and the credit taken for
' Containment closure, as opposed to Containment, integrity based on Generic Letter 88-17, the conse-quences are considered to be less. The analysis for a LOCA adequately covers the defense mechanisms in L0
i i:
[. . .
place to prevent a release of radioactive material
+
to the environment from a sustained loss of ND cooling at reduced NC System inventory.
Systems required to recover from a loss of ND are mainly located outside Containment and'administra-tively protected'to, ensure their availability for !
use during a loss of ND. Availability replaces )
operability with regard to safety system require- )
ments for responding to a loss of ND. Functionally
-capable systems can be in standby conditions,'under administrative. controls, or for use by the operator upon direction or need. The possibility of mal- j functions of equipment important to safety are_no different than those addressed in the FSAR. A variety of alternate means of core cooling is available to the operator to maintain the core in a y stable condition. -d I
The' margin of safety as defined in Technical Speci- -l fications is not reduced. T.S. 3.4.9.3 addresses overpressure protection for the NC System at low !
system temperatures. The basis states the one PZR j PORV has the capability to provide sufficient NC 1 pressure relief for one operating safety injection pump. At this point only one NV pump is used and if l two are required, an additional PORV can be opened.
Based on the information provided, an unreviewed safety question does not exist.
PT/2/A/4250/13C CA Pump #2 Head Verification: The purpose of this l procedure is to. perform a precision head-capacity. !
test on CA Pump #2. These results'will'be used for "
two purposes. The'first.is to detect possible pump . l degradation over time for various points on the head-curve. The second is to verify pump performance for j balancing CA flows to the steam generators per PT/2/A/4250/03E, CA System Discharge Control Valve j}
Throttling procedure. .The balance is different for various pump strengths.
.I In order to perform this test, CA Pump #2 will be !
aligned in recirculation to the~ Upper Surge Tank i
(UST) and data will be taken at various flowrates. q
, Turbine bearing lube oil cooler flow will be iso- l lated during this test. Precautions are taken to assure that temperature limits are not exceeded, and !
if these limits are approached, to valve the cooling l water back in until temperatures are reduced. This test will be performed in Modes 1, 2, and 3.
4 i
One CA Pump is allowed to be out'of'. service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per Tech Spec Action _ 3.7.1.2a. The two Motor-Driven CA Pumps will be available during this time. Therefore, neither the probability of nor consequences of an accident previously evaluated in the'FSAR will be increased. No new accident sce-narios will be created by the performance of this 1 test.
The pump will be run in a normal test-lineup with sufficient suction head verified prior to pump t start. The pump will also be run at flowrates well within design values. Therefore, neither the possibility of nor consequences of an equipment +
malfunction are increased. No new equipment mal- .
function possibilities are created.
Since the CA System will still have two available pumps and will be in a condition allowed by Techni-cal Specifications, the margin of safety will not be.
reduced. i OP/2/A/6250/01 Condensate and Feedwater System: This restricted change is being made to provide guidance for full D/P stroke verification of valves 2CA-149, 150, 151, and 152. These verifications will be done in Mode 5 , 1 and will not affect reactor safety. This change will not affect-any accidents previously evaluated i in the FSAR,.nor will the possibility of an accident r different than any evaluated in the FSAR be created.
Equipment safety will not be affected. CF pump speed will still be protected by an overspeed trip.
CF pump miniflow will still exist with PCF-13 (CF pump 2B Recirc Ctrl) in Auto. CF pump runout'i.s not expected. S/G overpressurization is not expected because vacuum will be lined up on each S/G being g tested. S/G overfilling should not occur because procedurally the Feed Reg and Feed Reg Byp valves will be closed at 70% Wide Range level. If closing the Feed Reg and Feed-Reg Byp valves does not stop water from going into the S/G, P-14:(High High S/G Level) will give a feedwater isolation signal tripping the operating CF pump. S/G chemistry will '
be maintained by being lined up to vacuum. Feed line water hammers are being minimized by opening the Feed Reg Byp valve prior to opening the Feed Reg valve.
The ND System will be in operation providing core cooling. The S/G will not be used as a heat sink "
for the NC System during this verification.
2_2
c l
=PT/2/A/4250/03D, Auxiliary Feedwater_ System Performance Periodic Change #12- Test: This procedure change is being written'to allow PT/2/A/4250/03D-testing to continue after 2SA2 i failure. Since 2SA2 circuitry and 2SA5 circuitry are parallel, the permissives necessary to' complete .
the test are given by 2SAS alone being open. 2SA2 is not required to be open to obtain the necessary permissives. <
During'the test block valves 2SA1 and 2SA4 are closed by prerequisites already in the procedure. -!
This test, as well:as all maintenance work involving 2SA2, will be completed prior to entering Mode 3, therefore, operability is not a concern in Modes 4, t 5, 6 and No Mode. .t Therefore, the probability / consequences of any-new or previously analyzed accident or equipment mal-function is not increased, and the margin of safety. .
is not reduced.
IP/0/A/3820/04 Operating Checkout of Limitorque and Rotork Valve Actuators, Change #239: This change incorporates NSM CN-20382. See the 50.59 evaluation for this- ,
NSM.
IP/2/A/3030/07H Maintenance and Functional Test Procedure for Main Steam Isolation Valve,-Change #2: This change incorporates NSM CN-50396. See the 50.59 evaluation for this NSM.
PT/2/A/4150/22, Total Core Reloading: In the Design Features Retype #1 Section Tech Spec 5.3.2, Control Rod Assemblies, is affected by a change that.is incorporated in this-retype. In this section it describes'that all control rods shall be clad with stainless steel '
tubing, this retype would allow the insertion of a fuel assembly into the core that contains an: RCCA clad with Inconel. ~The Tech Spec change. request was submitted April 6,1989, (see Attachment 'No.1). 'It is anticipated that NRC approval will be obtained by May 18, 1989. Peter LeRoy (G.0. Licensing) has received verbal approval from Kahtan .N. Jabbour, NRC, to proceed with Reactor Startup .if written approval is not received by May 18, 1989.
This modification'to the design of the RCCAs is '
exempt from the Nuclear Station Modification (NSM) process (see Section 3.4.1 of the Administrative D
n -
7 q
s r
r i
Policy Manual for Nuclear Stations and Section 1.5.1 of the Nuclear Station Modification Manual). Design ,
Engineering Department is responsible for reload o core design activities and maintains records of these activities in the Design Engineering Depart-ment-files.
This procedure is designed to load the core in a.
safe and orderly manner. This test is not described-in the FSAR. The closest description is found in FSAR Section 9.1.4 which describes the' process used in an incore fuel assembly / insert shuffle. FSAR Section 9.1.4 includes a description of'how the fuel .
is loaded from the SFP to the core. This procedure complies with the loading description found there.
It is different.in that the core is completely
. unloaded at the beginning, while FSAR Section 9.1.4 leaves fuel assemblies in the core at all times. 9 FSAR table 14.2.12-2 (page 2), Initial Fuel Loading Abstract describes the initial fuel loading proce-dure used at CNS. This procedure differs from the abstract in that temporary detectors are not used.
This revision includes the requirement that the Fuel Handling SR0 is to be present in the Reactor Build-ing whenever fuel is moved in the Reactor Vessel by non-licensed personnel. This is to comply with the latest version of'10CFR55.13b.
The fuel assembly RRN or insert ids will be identi-fied as the fuel assemblies are loaded into the core. Identifying fuel assemblies as they are loaded is required in FSAR Section 9.1.4-and 1s-mentioned in ANSI standard N15.8-1974. 'All fuel assemblies going into the core are verified to be in the proper locations and to contain the proper
~
inserts except 12P97K as part of PT/2/A/4550/09, Shuffle Verification, after the inserts are shuffled to the next cycle fuel assemblies. ' Hence verifying :
either the fuel assembly RRN or the insert'ID serves-to verify that the proper fuel assembly is being
[ loaded.
The loading sequence was revised to maximize-the number of fuel assembly " boxes" made from fresh and' once burned fuel assemblies to ease the loading of twice burned assemblies. And in addition to this there is a Limit and Precaution to avoid clusters of more than four new fuel assemblies. This is to ensure that Keff < 0.95 (Attachment #4).
Z4
r ; m.
Westinghouse has recently revised F-Specifications (F-4 and F-5, see Reference 2.12 of procedure) with Rev. #11. Rev. #10 has been compared to Rev. #11 ,
relative to this procedure:
i Attachment A to Appendix 1, " Typical Close Contact Movement of Fuel With Modified Grid Corners," has not changed from-Rev. #10. ,
Westinghouse did add to F-5 a new 6._4.8, ;
Diagonal Corner to Corner Contact (between fuel assemblies), as a result Enclosure 13.10'of this procedure, that offers'a suggested loading sequence, has been checked to ensure compliance
, with Rev. #11.-
A Note on BDMS OPERABILITY was added to clarify when the 80MS system would be INOPERABLE. This was based. r on'J. W. Caldwell's Memo to File #CN-214.04 on ENC operational characteristics dated 11-12-87.
The minimum water depth in the Reactor Vessel and the Spent Fuel Pool was changed to 95% and 37 feet respectively. A possible unreviewed safety question' ',
had been discovered earlier, in that the FSAR ,
requires 23' feet of. water over top of fuel assembly dropped on the Spent Fuel Pool floor while Tech Specs required 23 feet of water above the assemblies in the racks. This did not consider the possibility 4 of a fuel assembly being dropped on top of the Spent Fuel Pool racks. This question is being looked into by Design Engineering (see Attachment No. 5, a memo -
from Ken Ashe reply to PIR 0-C88-0342). It is not.
an unreviewed safety question; Design Engineering will: update the FSAR by' July 1, 1989.
No increase in the probability or consequences of an-accident evaluated in the FSAR is created by this procedure. This procedure creates no accident scenarios that are not already analyzed. The only ,
loads being moved are fuel assemblies. The accident discussed by FSAR.Section 15.7.4, Fuel Handling Accidents in the Containment and Spent Fuel storage Buildings, is bounding for fuel movement. The accident discussed in FSAR Chapter 15.4.7, Inadver-tent Loading and Operation of a Fuel Assembly in an-Improper Location, is also bounding for this proce-dure. To prevent this accident, the fuel assemblies are identified as they -are put into the core (see above) and PT/2/A/4550/03C, Post-Refueling Core !
Loading Verification, is performed after all fuel a assemblies are loaded. PT/2/A/4550/03C also veri-fies that fuel assemblies designated for RCCAs have an RCCA as their core component.
i L
. ~
-i
. t
+ ,
Inadvertently a rodlet from 12P97K (a burnable poison assembly, 'WABA') was damaged during han- ,
dling. Due to time constraints, the entire poison assembly (12P97K) will be replaced by 12P105K. The
- 12P97K and 12P105K are functionally identical-and will be both manufactured by Westinghouse. Since PT/2/A/4550/09,- Shuffle Verification, has already a been completed for Cycle 3 reloading i PT/2/A/4550/03C, Post-Refueling Core Loading Veri- 4; 1
fication, will require a restrictive change to specifically verify that-12P105K is inserted in Fuel 4 Assembly _Q54 in Core location B-9. No unroviewed safety. question is resulted from replacing 12P97K with 12P105K.
This procedure does not require off-normal operation of safety equipment so there will be no increase in.
the probability or consequence of a malfunction of ;
equipment important to safety _as previously'evalu-ated in the FSAR nor create a possibility of equip-ment malfunction important-to safety different than any evaluated in the FSAR.
No safety features are jeopardized and no Tech Spec.
requirements for refueling are violated by this t
procedure, therefore, the margin of safety is not reduced, pT/1/B/4200/66, Rev. O Reactor Makeup Water Storage Tank Recirculation Time Test: This test procedure is designed to periodi-cally check the recirculation time of the Reactor Makeup Water Storage Tank (RMWST) to determine if the recirculation time specified _in-the NM ' system operating procedure is adequate. It involves injecting a lithium hydroxide solution into the RMWST to obtain a predetermined lithium concentra-tion, recirculating the tank with the reactor makeup water pump, and taking samples at regular' intervals.
The recirculation time is determined by the. amount of time it takes for the lithium concentration to stabilize at the predetermined concentration.
Lithium is always present in the reactor coolant system and any lithium in the RMWST will be recycled ;
to the reactor coolant system during unit operation, Therefore, the probability or consequences of an accident that has been previously evaluated in the FSAR will not be increased, nor will the possibility of an accident that has not been evaluated in the e FSAR be created.
26
p-h-
L ,
- 0.2 ppm lithium hydroxide in a domineralized water i solution will have a conductivity value of 6,75 )
(? pmhos/cm and a pH value of 9.5. This is above the ,
conductivity' specification of 2.0 pmhos/cm for the '
U RMWST (the RMWST should be full of demineralized water) listed in. Table ~9.3.5-2 of the FSAR. This-
,_ high conductivity value indicates that an ionic contaminant is present in the RMWST. The lithium hydroxide contaminant is intentionally being added .,
, to the RMWST to determine the minimum recirculation time.
The lithium hydroxide solution will not cause.
problems in the' RMWST or the reactor coolant system.- .
Even the high value of 9.5 pH will not harm any stainless steel piping, pH values of 9.5 have-lower '
stainless steel corrosion relates than lower pH values. If the 0.2 ppm lithium hydroxide solution- ,
were added to the reactor coolant system, it would cause the lithium concentration that already exists in the reactor coolant system to decrease at a P slower rate than it would if it were being diluted with deaerated demineralized water. Therefore, this is not a problem. .
(
All permanent plant equipment mentioned in this procedure is operated per previously evaluated and approved procedures. Also, the pump used to inject the lithium hydroxide solution is not capable of producing a discharge pressure over 100 psi. The ,
low maximum pump discharge pressure removes the possibility of this test overpressurizing any piping. Therefore, the probability or consequences of malfunction of equipment important to safety that has been previously evaluated in the FSAR will not ,
be increased, nor will the possibility of malfunc-tion of equipment important to safety not previously evaluated be created.
No Technical Specifications are affected by this !
procedure.
IP/1/A/3160/01, Annulus Ventilation System: This procedure change Change #26 ensures Train A of the Annulus Ventilation (VE) ,
System Pressure Controlled dampers operate as designed. Therefore, the Technical Specifications are not affected by this procedure change.
1 ZZ
h ,<
g f
l This procedure change adds provisions to allow '
cycling dampers 1AVS-D-2 and 1AVS-D-3 (Train A Annulus Supply, and Train A Unit Vent Supply ~, i respectively). The VE System is designed to 1) produce and maintain a negative pressure in the annulus following a~LOCA, 2) minimize radioisotope.
release following a LOCA, and 3) provide long term -
fission product removal by decay and filtration. A
-safety injection signal actuates these dampers, while the controls involved in this procedure change g modulate as required to achieve and maintain the- ,
annulus pressure at a negative pressure. As a
, radiation protection design feature, the air re-' j leased from-the annulus is passed through the filter 1
, units, through damper 1AVS-D-3, and exhausted through the Unit Vent.
Since the VE System is provided with two independent 5 100% capacity filter trains, and since the:syst'em- :
meets the single failure criteria, this procedure >
change does not prevent the system from performing-its required safety function. Assuming 1AVS-D-3 is in the open position during calibration coincident with a LOCA, the annulus pressure will approach the desired negative pressure region as required by the system design. Assuming this same damper is closed, the Train B portion will operate to bring the ;
annulus to_the required pressure, Also, during both '
of these cases, any releases through the Unit Vent are filtered to minimize radiation releases.
The consequences or probability of an accident evaluated in the FSAR is not increased or created by thi.s change. Also, the probability of a malfunction of equipment previously evaluated in the FSAR is not ,
increased or crt;ted. The consequences of a mal- ,
function of equipment impcrtant to safety evaluated in the FSAR is not affected based upon the single- :
failure criteria. The margin _ of safety defined in the basis of the Technical Specifications.for Catawba is not reduced. Based upon this evaluation, ,
an'unreviewed safety question does not exist.
Therefore, the FSAR is not affected and an s unreviewed safety question does not exist as a result of this procedure change.
PT/2/A/4150/18 Fuel Assembly / Insert Shuffle procedure: This.
procedure is designed to provide a safe transfer of fuel assembly inserts from the Cycle 2 fuel assem-blies to their Cycle 3 fuel assemblies. Only inserts are moved in this procedure, not any fuel l
lm, , ,
[;[
assemblies. This test is not described in the FSAR, but-FSAR Section 9.1.4 describes an in-core shuffle '
general description which this procedure follows.
,The difference is that this procedure takes place entirely in the Spent Fuel Pool.
The general procedure is for the fuel handling personnel to move the appropriate insert ' handling tool over the specified insert, remove the insert ;
and place it in its final fuel assembly. Indepen - ; '
dent Verification is required both for the initial position of the insert'and the final position.of.the
't' insert.
The Cycle 2 to Cycle 3 Shuffle is different than-encountered before-in that there is a possibility that'B&W Fuel Company Demonstration RCCAs might be inserted into the Catawba Two Cycle 3 Core. The s procedure has been adjusted to be flexible enough to.
' handle the possibility of inserting the B&W RCCAs.
In Section 8.0 there is a step to verify with Design Engineering that the Core Loading Plan ~(CNM-2201.30-0002-001) is the correct pattern for Cycle 3. If it .is determined that these B&W RCCAs will go.in the next !
cycle, then Tech Spec 5.3.2 will need to be changed to allow a single RCCA to be cladded in Inconel instead of Stainless Steel in which case the 50.59 Evaluation for PT/2/A/4150/22, Total Core Reloading, will address this. concern.
Fuel assembly RRNs and insert ids are not checked by -t '
this procedure. It is designed to move inserts based only on the positions in the Spent Fuel Pool.
However, PT/2/A/4_550/09, Shuffle Verification, is required to be performed after the shuffle is complete. This test will verify that each fuel ,
assembly that-is going back into the core contains the correct insert for Cycle 3 and is-in the proper 1 Spent Fuel Pool location. In addition, PT/2/A/4550/030, Post Reloading Core Verification will be performed af ter . fuel loading to verify that the fuel ~ assemblies are in the proper core location for Cycle-3,.and I that all' fuel assemblies that should contain RCCAs 1 L contain RCCAs.
L This procedure is bounded by the accident analysis performed by FSAR Chapter 15.4.7, Inadvertent Loading and Operation of a Fuel Assembly in an Improper Location. This chapter also analyzes .
operation and loading of a fuel assembly with an incorrect insert.
This procedure requires movement of inserts and fuel handling tools over the spent fuel in the Spent Fuel 29
b:
'r Pool, but the inserts'and handling tools weigh less j than a~ fuel assembly. Therefore, the accident a analysis in FSAR Chapter 15.7.4, Fuel-Handling Accidents, is bounding.- In addition, precautions are taken in the procedure to ensure-that objects are not dropped into-the' Spent Fue1~ Pool.
The latest philosophy on Thimble Plugs (from Paul Bailey D.E.) is that if the thimble plug rods are-
~ bent more than 1 inch, then the thimble plug should not be bent back into alignment and reinserted into-a Fuel Assembly going back into the. core due to the possibility of the rod breaking off during operation causing a) loss part and, b) allowing bypass flow-though F/A guide thimble. As a response to this latest concern, this procedure will not reinsert any bent Thimble Plugs into F/As.
This procedure does not require off-normal operation of safety equipment so there will be no increase .in the probability or consequences of a malfunction of-equipment important to safety as previously evalu-ated in the FSAR nor create-a possibility of equip-ment malfunction important to safety different than any evaluated in the FSAR.
No safety features are jeopardized and no Tech Spec requirements are violated by this procedure, there-fore, the margin of safety is not reduced.
EP/2/A/5000/10, Steam Line Break Outside Containment: The
' Retype #4 significant change to this procedure is a deletion of Step 2.b which was a redundant check for a faulted S/G. The RNO of-this step was used to check for an isolated steam line break downstream of the MSIVs. EP/01, reactor trip or-safety injection,;has. >
been changed under retype #5 so that the flowpath '
for terminating S/I after an isolated steam line break'is no longer through this procedure, but through EP/18, S/I termination following spurious S/I. It is felt that an isolated steam:line break can be treated like a spurious S/I since the cooldown is quickly terminated due to isolation of the break by main steam-isolation. This is different from an unisolated steam line break where-an extended cooldown exists due to the complete 1 blowdown of the faulted S/G. See EP/01 retype #5 10CFR 50.59. evaluation for justification of this change.
Since the flowpath for responding to an isolated steam itne break has changed, this deleted step is no longer needed in this procedure since it only applies for the isolated steam line break scenario.
This change will require a change to the Catawba 30 1
EPGs -Design Engineering (Safety Analysis Group):
has been notified and agrees with_ this change and is pursuing the change to;the Catawba EPGs. Since this change does not affect the evaluation of steam line-breaks as given in the FSAR nor does it create the possibility of an accident different from that evaluated in the FSAR, it is felt that this change does not represent an unreviewed safety question..
EP/1/A/5000/1A-, The following changes are included in' Retype #14:
Retype #14 1)- Changed Step #15.a RNO- from " Notify IAE to initiate SWR 3994" to " initiate SWR 3994".
Operations has copies of this SWR at the SR0 desk and it is not necessary to have IAE issue the work request.
2)' Changed Step #25.a.2) RNO from "...Then notify IAE to initiate SWR..." to "...then initiate SW R . . . " . Operations has copies of this SWR at the SR0 desk and it is not necessary to have IAE issue the work request.
- 3) Changed step reference to EP/01 on Enclosure 1.
This was done to reflect changes in EP/01 under retype #8.
- 4) Corrected various format errors to ensure compliance with the EP Writer's Guide.
None of the-'above listed changes affect the proce-dure in a significant manner or alter its intent.
Therefore, the changes listed above do not represent an unreviewed safety question.
- 5) One significant change to this procedure is_the deletion of the requirement to verify proper
-operation of P-4 contacts following every Rx trip. This requirement was added to the-procedure due to a commitment made by Duke Power Company (for CNS) in a letter from H. B.
Tucker to D.G. Eisenhut of the NRC, dated December 31, 1984, which provided additional information on Duke's response to Generic Letter 83-28, Item 3.1.2, review of vendor engineering recommendations for reactor trip system components. In this letter, CNS com-mitted to verifying proper P-4 operation after every Rx trip as a means of satisfying the concerns brought about by Westinghouse in data letter 79-4.
Westinghouse data letter 79-4 addressed the potential for an undetected failure of the P-4 permissive (an ESF actuation system). The data 8_1
+
t letter stated "Since there are no tests to verify proper operation of P-4 contacts in the reactor trip switchgear, it is recommended that-the correct function of P-4-permissive ~is '
checked following a reactor trip due to actua--
tion of safety injection".
When this commitment.was made,' Catawba did not test P-4 operation as a part of SSPS testing.
However, the proper operation of the p-4 permissive is performed by Catawba IAE person-nel bi-monthly in accordance with .'
IP/1(2)/A/3200/08A(B), Train A(B) reactor trip breaker trip actuating device functional and operational test. .This; procedure tests. proper F operation of the P-4 contacts using both the Rx trip breakers and the Rx trip bypass breakers.
Since Catawba has procedures which will test c
-P-4 operation bi-monthly, it is felt that the -
concerns of Westinghouse data letter 79-4 are adequately covered by this testing and that it' is no longer necessary to continue checking i proper P-4 status after every Rx trip'.- Catawba prefers to eliminate this verification after each Rx trip since this diverts limited man-power which should be directed towards other vital tasks.
In response to.this change, the NRC'is being notified, by letter from H. B. Tucker, of our }
intent to stop the verification of. proper P-4 jl contact operation- following every Rx trip. -
Since the concerns of Westinghouse data letter 1
.79-4 are being addressed at CNS by IAE periodic:
testing, it is felt that this change will not-
. increase the probability or possibility of an accident (or malfunction of equipment) previ-ously evaluated (or different from any accident previously evaluated) in the FSAR. Therefore, this change does not represent an unreviewed i safety question.
EP/2/A/5000/1A, Reactor Trip Response: The following changes are Retype #8 included in retype #8,
- 1) Reworded Step #2 for human factors concerns,
- 2) Changed Step #3 to provide options in aligning a steam supply to the as header. The intent of this step is to ensure a steam supply is '
< available for the as header following a Rx trip. Since the as header can be supplied from 3 different sources, the step was expanded to take advantage of the options available.
m p ,
e i:
p
- 3) Modified Step #6.a. RNO to include information on how to place ORPI-in half accuracy mode. ,
- 4) Changed Step #14.a. RNO from " Notify IAE to '
initiate SWR 4118" to " initiate SWR 4118".
Operations has copies of this SWR at the SRO desk and it is not necessary to have IAE issue !
F the work request.
5). Changed Step #24.9.2) RNO-from "...Then notify IAE to initiate SWR..." to "...Then initiate g SWR...". Operations has copies of this SWR at 1 N the SRO desk and it is not necessary to have r IAE issue the work request.
- 6) Changed step reference to EP/01 on Enclosure 1, This was done to reflect changes in' EP/01 under .
retype #5.
- 7) Modified order of natural circulation parame- l ters on Enclosure 1 to address human factors a concerns. 1 None of the changes listed above affect the procedure in a significant manner or alter its !
intent. Therefore, the~ changes listed above do not represent an unreviewed safety question,
- 8) One significant change to this procedure is the '
deletion of the requirement to verify proper operation of'P-4 contacts following every Rx trip. This requirement was added to the procedure due to a commitment made by Duke Power Company-(for CNS) in a letter from H. B.
Tucker to D. G. Eisenhut of the NRC, dated >
December 31, 1984, which provided additional -
information on Duke's response to Generic 1 Letter 83-28,. Item 3.1.2, Review of Vendor '
Engineering Recommendations for Reactor Trip
. System Components.- -In this letter, CNS com-mitted to verifying proper P-4 operation after every Rx trip as a means of satisfying the concerns brought about by Westinghouse in data letter 79-4, Westinghouse data letter 79-4' addressed'the potential for an undetectable failure of the P-4 permissive (an ESF actuation system). The data letter stated "Since there are no tests to
, verify proper operation of P-4 contacts in the L reactor trip switchgear, it is recommended that the correct function of P-4 permissive is checked following a reactor trip due to actua-tion of safety injection".
_83
y .
E When this commitment was made, Catawba did not test P-4 operation as a part of SSPS testing,
, However, the proper geration of the P-4 permissive is performed by Catawba IAE person-nel bi-monthly in accordance with IP/1(2)/A/3200/08A(B), Train A(B) reactor trip breaker trip actuating device functional and operational test. This procedure tests proper l
cperation of the P-4 contacts using both the Rx
, trip breakers and the Rx trip bypass breakers.
Since Catawba has procedures which will test
,~ P-4 operation bi-monthly, it is felt that the concerns of Westinghouse data letter 79-4 are
, adequately covered by this testing and that it l- is no longer necessary to continue checking proper P-4 status af ter every Rx trip. Catawba
- - prefers to eliminate this verification after each Rx trip since this diverts the limited manpower which should be directed towards other vital tasks. In response to this change, the NRC is being notified, by letter from H. B.
Tucker, of our intent to stop the verification of proper P-4 contact operation following every Rx trip. Since the concerns of Westinghouse data letter 79-4 are being addressed at CNS by IAE periodic testing, it is felt that this change will not increase the probability or possibility of an accident (or malfunction of equipment) previously evaluated (or different from any accidvd previously evaluated) in the FSAR. Therefe m this change does not repre-sent an unreviewed safety question.
EP/1/A/5000/01 Reactor Trip or Safety Injection:' The following EP/2/A/5000/01 changes have been made in retype #8:
- 1) Various steps in thts procedure have been rewritten, reformatted and/or relocated within this procedure to address certain human factors concerns. The changes have been made to make the steps easier to understand and communicate. l However, even though these steps have been l modified, the intent of each step-has not been !
changed. The only change is the manner in which the information is presented. Design L Engineering has reviewed and approved all of the changes mentioned above and feel that the intent of the orocedure has not been changed. .
Therefore, it is felt that these changes do not {
represent an unreviewed safety question, i
- 2) Deleted actions to open MG-set feeder breakers on ILXC and ILXD under Step #1.b.RNO. These ;
actions are listed in EP/2A1 and will be i
84
i- , ,
L ;
r 4 accomplished per that procedure should the l actions in EP/01 be unsuccessful in tripping !
!. the reactor. Since these actions are given in ,
EP/2A1, it is not necessary that they be listed I in this procedure. Therefore, since the intent a l of this step is still being carried ~ out, it is felt that this change does not represent an p unreviewed safety question.
- 3) Deleted reference within the procedure for checking VX system operation after an Sp signal j '- and 9 minute time delay. This action is 1 already covered under the step which checks proper monitor light status after an Sp signal. <
, This step is also covered by Enclosure 1 of ;
EP/10. Since this step is adequately covered :
by these steps, it is not necessary that it be checked by a separate step in EP/01. There-
, fore, it is felt that this change does not L represent an unreviewed safety question. [
- 4) Changed RNO actions in Step #2 which is to -
verify that a turbine trip has occurred. 4 Changed second action to manually initiate main steam isolation if a manual turbine trip is '
unsuccessful. Validation on the simulator has L shown that a turbine trip failing to occur !
following a reactor trip will result in an automatic main steam isolation within 5 sec- '
onds. This change is directing the operator to.
perform those actions which will occur auto-matica11y within a few seconds. The purpose of this change is to reduce the-time-spent in Ep/01 trying to attain a turbine trip when.more <
significant abnormalities may be occurring, i Immediate actions in EPs are often designed to verify or complete automatic protective ac-tions. Applying one assured corrective action.
instead of several possibic corrective actions .
( permits a more expeditious approach to diag- i nostics and subsequent recovery options.
FSAR Sections 15.2.2 through 15.2.4 discuss the !
analyses of a loss of external load, turbine generator trip and inadvertent closure of main af ,'
steam isolation valves, respectively. The closure of the main steam isolation valves ,
results in a turbine trip, thus the analysis is :
the same for the two events. Loss of external
, load is considered less severe than a turbine "
trip due to the slower reaction tire for stop i valve closure. The turbine trip analysis i assumes nominal power, pressure and temperature prior to the trip. The reactor does not trip :j I
il l'<, l l
l coincident with the turbine, but trips on f i
resultant high. pressurizer pressure, level or f cm overtemperature Delta T. This delay in reactor ;
shutdosn results in rapidly increasing primary /- 4 secondary system pressures. The pressure transient in both systems is maintained within design parameters by system related safety valve actuation. In EP/01 a reactor trip has been verified prior to main steam isolation, so the resultant pressura transient is within the bounds of the turbine trip analysis. The plant. ;
is stabilized after the turbine trip malfunc- ,
tion in EP/1A, reactor trip response. Heat ,
removal for the primary system is accomplished '
using the S/G PORVs in manual and auxiliary feedwater. ;
Since a turbine trip and an inadvertent closure [
of the main steam isolation valves are both >-
-accounted for under the same analysis, then the activation of main steam isolation to assure a ,
stoppage of steam flow to the turbine does not !
increase the probability of an accident previ-ously analyzed in the FSAR.
The consequences of a main steam isolation are different than a normal turbine trip, because steam dump availability is denied. However, adequate steam relief capability-is provided for each steam line by the main steam line safety valves. In addition, operator control !
of the S/G p0RVs can be established without-resetting the main steam isolation signal.
Therefore, a failure of_ the reset capability of this protective feature does not impede the :
operator from stabilizing the plant after steam .
flow to the turbine is isolated. .Since both the turbine trip and main steam isolation valve s closure events are covered by the same analy-4 sis, no new accident scenario has been created.
The analysis does not take credit for steam dump actuation or S/G PORV operation to relieve-secondary pressure. Neither does it take .
credit for S/G feed from the main feedwater pumps or the auxiliary feedwater system. Based-on these conditions of the analysis, neither the probability nor the consequences of a ,
malfunction of equipment important to safety are increased. Initiating a main steam isola-tion to isolate steam flow does increase the ,
possibility of primary and secondary systed. '
safety valve operation. However, the failure of each of these system's safety valves has been previously evaluated in the FSAR (primary -
15.6.1, secondary 15.1.3).
81
l The margin of safety provided by Tech Spec 3.3.4, turbine overspeed protection, is not r affected by this change. The operator response time to an inoperable turbine. protective function whether overspeed protection or other, during an actual event requiring a trip, is '
shortened by this change, the manual isolation of steam to the turbine assures turbine pro-tection, while not jeopardizing reactor safety .
or operator attention to more significant l transients, f
- 5) Deletion of the requirement to verify proper operation of P-4 contacts following every reactor trip. This requirement was added to the procedure due to a commitment made by Duke Power Company (for Catawba) in a letter from H.
B. Tucker to D. G. Eisenhut of the NRC, dated l December 31, 1984, which provided additional information on Duke's response to Generic l Letter 83-28, Item 3.1.2, Review of Vendor l Engineering Recommendations for Reactor Trip ;
System Components. In this letter, Duke .
committed to verifying proper P-4 operation '
after every Rx trip as a means of satisfying the concerns brought forth by Westinghouse in data letter 79-4.
Westinghouse data letter 79-4 addressed the !
potential for an undetectable failure of the P-4 permissive (an ESF actuation system). The data letter stated "Since there are no tests to verify proper operation of P-4 contacts in the reactor trip switchgear, it is recommended that the correct function of P-4 permissive is '
checked follow'ng a reat. tor trip due to actu-ation of safety injection".
When this commitment was made, Catawba did not test proper P-4 operation as a part of SSPS testing. However,'the proper operation of the.
P-4 permissive is performed by Catawba IAE -!
personnel bi-monthly in accordance with IP/1(2)/A/3200/08A(B), Train A(B) reactor trip i breaker trip actuating device functional and operational test. This procedure tests proper operttion of the P-4 contacts using both the Rx trip breakers and the reactor trip bypass breakers. Since Catawba has procedures which will test P-4 operation bi-monthly, it is felt that the con: erns of Westinghouse data letter 79-4 are adequately covered by this testing and that it is no longer necessary to continue i checking proper P-4 status af ter every Rx trip.
81
Catawba prefers to eliminate this verification-after each Rx trip since it diverts limited manpower which should be directed towards other tasks. In response to this change, the NRC is being notified, by letter frem H. B. Tucker, of our intent to stop the verification of proper P-4 contact operation following every Rx trip.
Since the concerns of Westinghouse data letter ,
79-4 are being addressed at CN5 by IAE periodic testing, it is felt that this change will not increase the probability or possibility of an accident (or malfunction of equipment) previ-ously evaluated (or different from any accident previously evaluated) in the FSAR. Therefore, this change does not represent an unreviewed safety question.
- 6) Changed flowpath within EP/01 for responding to an isolated steam line break downstream of the .
MSIVs. Originally, EP/10 would direct the operator to perform EP/10, steam line break -
outside containment, if an isolated steam line break was the event. This in turn would lead ,
to EP/IDI, S/I termination following steam line ,
break, to terminate S/I. This change directs r the operator to continue through Ep/01 until reaching the S/I termination criteria. If the-criteria can be met, the operator would now perform EP/10, S/I termination following spurious S/I, to terminate S/I. It is felt that an isolated steam line break can be treated like a spurious S/I since the cooldown is quickly isolated by main steam isolation. ,
This is different from an unisolated steam line i break outside containment where an extended cooldown exists due to the blowdown of the faulted S/C. The steps which are performed in EP/1D are the necessary actions to isolate flow to and from a faulted S/G. These steps don't apply to an isolated steam line break since main steam isolation terminates the leak. The '
cooldown which occurs on an unisolated steam line break raises the potential that shutdown margin could be lost. This is not a real concern with an isolated steam line break since the cooldown is quickly terminated by main steam isolation. Therefore, it is felt that the quickest and most reliable path for re-
'l sponding to an isolated steam line break is through EP/01 and EP/18. Design Engineering has been notified and agrees with this change.
This change will require a change to the Catawba EPGs. Design Engineering (Safety Analysis Group) is pursuing making this change.
B8
y fr; Since this change does not affect the evalua-tion of steam line breaks as given in the FSAR, nor does it create the possibility of an accident different from that evaluated in the FSAR, it is felt that this change does not represent an unreviewed safety question.
AP/1/A/5500/02 Turbine Generator. Trip: The changes to EP/01 AP/2/A/5500/02 Reactor Trip or Safety Injection and AP/02 Turbine Generator Trip consist of reducing the alternate methods of tripping the Main Turbine if the auto-matic trip function does not work. The change to EP/01 directs the operator to manually trip the Turbine. If the Turbine does not trip, then the operator manually initiates Main Steam Isolation.
The change to AP/02 directs the operator to manually trip the Turbine. If the Turbine does not trip, then the operator manually trips the Reactor and initiates Main Steam Isolation. Former guidance allowed several other operator actions to be.per-formed prior to closing the MSIV and MSIV bypass valves to stop steam flow to the Turbine.
The purpose of the change is to reduce the time spent in EP/01 trying to attain a turbine trip when more significant abnormalities may be occurring.
Immediate actions in EP/APs are often designed to verify or complete automatic protective actions.
Applying one assured corrective action instead.of several possible corrective actions permits a more expeditious approach to diagnostics and subsequent recovery options.
FSAR Sections 15.2.2 through 15.2.4 discuss the analyses of a loss of External Load, Turbine Gener-ator Trip and Inadvertent Closure of Main Steam Isolation valves,.respectively. The closure of the
, Main Steam Isolation valves results in a Turbine Trip, thus the analysir is the same for the two.
incidents. Loss of External Load is considered less severe than a Turbine trip due to the slower reac-tion time-for stop valve closure. The Turbine Trip analysis assumes nominal power, pressure and tem-perature prior to the trip. The Reactor does not trip coincident with the Turbine, but trips on .
resultant high pressurizer pressure, level or overtemperature delta T. This delay in Reactor shutdown results in rapidly increasing prima-ry/ secondary system pressures. The pressure tran-sient in both-systems is maintained within design !
parameters by system related safety valve actuation.
In EP/01 a reactor trip has been verified prior to h Main Steam Isolation, so the resultant pressure 1 l
~
L b
n.
transient is within the bounds of the Turbine Trip analysis. For AP/02 the operator is directed to manually trip the Reactor prior to initiating Main Steam Isolation. Again the resultant pressure transient is within the bounds of the Turbine Trip analysis. The plant is stabilized after a Turbine trip malfunction in EP/1A Reactor Trip Response.
Heat removal for the primary system is accomplished using the S/G PORVs in manual and the auxiliary feedwater.
Since a Turbine trip and an inadvertent closure of !
the Main Steam Isolation valves are both accounted for under the same analysis, then the activation of a Main Steam Isolation to assure a stoppage'of steam flow to the Turbine does not increase the probabil-ity of an accident previously analyzed in the FSAR. -
The consequerces of a Main Steam Isolation are ,
different than a normal Turbine trip, because stee.m dump availability is denied. However, adequate steam relief capability is provided for each steam line by the main steam line safety valves. In addition, operator control of the S/G PORVs can be established without resetting the Main Steam Isolation signal. Therefore, a failure of +.he reset capability of this protective feature does not impede the operator from stabilizing the plant after steam flow to the Turbine is isolated. Since both Turbine trip and Main Steam Isolation valve closure events are covered by the same analysis, no new i accident scenario has been created.
The analysis does not take credit for steam dump actuation or S/G PORV operation to relieve secondary pressure. Neither does it take credit for S/G feed from the main feedwater pumps or the auxiliary '
feedwater system. Based on these conditions of the analysis, neither the probability nor the conse-quences of a malfunction of equipment important to safety are increased.
Initiating a Main Steam Isolation to isolate steam flow does increase the possibility of primary and secondary system safety valve operation. However, the failure of each of these system's safety valves has been previously evaluated in the FSAR. (Primary
- 15. 6.1, Secondary 15.1. 3) .
The margin of safety provided by Technical Specifi-cation 3.3.4 Turbine Overspeed Protection is not affected by this change. The operator response time to an inoperable Turbine protective function whether overspeed protection or other, during an actual
y.I-event requiring a trip, is shortened by this change.
The manual isolation of steam to the Turbine assures Turbine protection, while not jeopardizing reactor safety or operator attention to more significant transients.
PT/2/A/4150/17 Total Core Unloading: This procedure is designed to unload the core in a safe and orderly manner. This test is not described in the FSAR, The closest description is found in FSAR Section 9.1.4 which describes the process used in an incore fuel assem-L bly/ insert shuffle. FSAR Section 9.1.4 includes a description of how the fuel is off-loaded from the core to the SFP.- This procedure complies with the off-load description found there. It is different in that it completely unloads the core while FSAR Section 9.1.4 leaves fuel assemblies in the core at all times. ,
F This revision includes statements requiring the Fuel Handling SRO to be present in the Reactor Building whenever fuel is moved in the Reactor Vessel by non-licensed personnel. This is to comply with the latest version of 10CFR55.13.
The unload sequence was revised to both maximize the number of open water off-indexing of fuel (i.e. to move an assembly to open water before lifting out of-the core) and to remove secondary source assemblies in a manner recommended in Westinghouse internal letter from R. A. Kerr to C. Gerstberger dated 10-20-83. The removal of secondary sources is designed to ensure that coupling between source, fuel and detectors is maintained. These changes do not affect safety in a negative manner.
A Note on BDMS OPERABILITY was added to clarify when the BDMS system would be INOPERABLE. This was based on J. W. Caldwell's Memo to File #CN-214.04 on ENC operational characteristics dated 11-12-87.
The minimum water depth in the Reactor Vessel and SFP was changed to 95% and 37 feet respectively. A possible unreviewed safety question had been dis-covered earlier, in that the FSAR requires 23 feet of water over top of fuel assembly dropped on the SFP floor while Tech Specs require 23 feet of water 1 of assemblies in the rack. This did not consider the possibility of a fuel assembly being dropped on .
to the top of the SFP racks. This question is being looked into by Design Engineering (see the attached l memo from Ken Ashe reply to PIR 0-C88-0342). It is "
not an unreviewed safety question; Design Engineer- )
ing will update the FSAR by July 1, 1989.
l 91 j.
y 5
L t :
i No increase in the probability or consequences of an accident evaluated in the FSAR is created by this procedure. Thit procedure creates no accident :
[ scenarios that are not already analyzed. The only -
load being moved is a fuel assembly. The accident discussed by FSAR Section 15.7.4, Fuel Handling :
Accidents in the Containment and Spent Fuel Storage Buildings, is bounding. r This procedure does not require off-normal operation i of safety equipment so there will be no increase in the probability or consequence of a malfunction of equipment important to safety as previously evalu-
, ated in the FSAR nor create a possibility of equip-ment malfunction important to safety different than any evaluated in the FSAR.
No safety features are jeopardized and no Tech Spec '
requirements for refueling are violated by this '
procedure, therefore, the margin of safety is not reduced.
OP/0/A/6100/06, Reactivity Balance Calculation: Retype #13 of Retype #13 OP/0/A/6100/06 is a major reformatting of the procedure. The retype performs various calculations 1 related to maintaining core reactivity within Tech Spec limits; this achievement will necessarily keep the core's reactivity within the assumptions of various FSAR accidents assumptions.
The retype is arranged into'6 enclosures. The ,
individual enclosures' discussions are below: !
Enclosure 4.1: This enclosure calculates the ,
estimated critical NCS boron concentration (ECB) :
based on other assumed core reactivity conditions (i.e. core burnup, desired critical RCCA position, core xenon worth, and core samarium worth). The calculated ECB is then used in either the Control-ling Procedure for Unit Startup or Fast Recovery OP as the boron to which the NCS is adjusted in prepa- '
ration for achieving criticality. There is a
" caution" before choosing the desired critical RCCA position to prevent the critical RCCA position from being either below the RCCA insertion limits n above any temporary RCCA withdrawal limits; this prevents the NCS boron from being adjusted to a ,i concentration which results in either a loss of adequate shutdown margin g an MTC in violation of Tech Specs when criticality is achieved.
- Enclosure 4.2: The first part of this enclosure calculates the estimated critical RCCA position (ECP) based on other assumed core reactivity R
n conditions (i.e. core burnup, NCS boron concentra-tion, core xenon worth, and core samarium worth).
Once the ECB of Enclosure 4.1 is
- achieved per Controlling Procedure for Unit Startup or Fast Recovery OP, this ECP is calculated with the actu-ally achieved NCS boron to obtain the ECP and 2 L " windows" about this RCCA position. This ECP satisfies TS Surveillance 4.1.1.1.1.c and demon-strates that the ECP will be below any temporary RCCA withdrawal limits. The first " window" about the ECp is 1 500 pcm, and defines the " expected" result of the estimated critical RCCA position calculation. The second " window" about the ECP is i 1000 pcm, and defines the " acceptable" result of the estimated critical RCCA position calculation. If during the approach to critical, criticality is indicated outside of this second window, the startup will be halted and an evaluation done befo're criti-cality is attempted again under the guidance of Reactor Engineering. This value of + 1000 pcm off ECP is in the spirit of the defined W reactivity anomaly" of TS Surveillance 4.1.1.1.2.
The tecond part of this enclosure is used during the l approach to criticality. It performs I/Ms as control banks are withdrawn to extrapolate a pre-dicted critical RCCA position'(PCP). This predic-tion is used with the various constraints on criti-cal RCCA position (i.e. RCCA insertion limits, withdrawal limits, and i 1000 pcm about ECP) to ensure core reactivity is behaving acceptably and as expected as criticality is achieved.
Enclosure 4.3: This enclosure calculates a conser-vative value for current shutdown margin when there are inoperable RCCAs at power. This is required by TS Action 3.1.3.1.a and C 3. The calculation is ,
conservative in that it uses the lowest total RCCA worth for a given fuel cycle (from the Data Book) {
and the highest, highest stuck RCCA worth for a given fuel cycle. Also,-~the worst case power defect is used to account for larger-than-nominal power defects which could occur dut-ing AFD oscillations.
Also, the penalty assumed for the inoperable RCCAs is taken as the larger of BOL and EOL calculated values for a given fuel cycle, j Enclosure 4.4: This enclosure calculates the required NCS boron to ensure shutdown margin when credit for xenon worth cannot be taken (T-AVG <
500 F). This satisfies TS 3.1.1.1 or 3.1.1.2. TFe !
required NCS botons as a function of T-AVG and core burnup are provided in the Data Book. These values b
E 1:
1
assume no xenon and equilibrium samarium and the highest worth RCCA stuck out (as required by the TS definit?on of shutdown margin); thus the considera-tions of TS Surveillance 4.1.1.1.1.e and 4.1.1.2.b are addressed. (There is a potential for samarium worth to be very slightly less than equilibrium,
- t. The amount less than equilibrium is at most + worth 50 pcm or
- 5 ppm.. This is within the accuracy of boron measurement and is considered insignificant.)
Enclosure 4.5: The first part of this enclosure r
< calculates the combination of NCS boron and xenon
. and samarium worths to ensure shutdown margin (when T-AVG > 500'F). This satisfies TS 3.1.1.1. The required NCS boron is obtained in the same manner as in Enclos"re 4.4. The calculated values of xenon and samarium worths are used (reduced by a conser-vative 10%) to supplement NCS boron to demonstrate adequate shutdown margin. The affect of this part of the enclosure is to show for what times shutdown margin is met given current NCS boron and xe-non/semarium behavior.
The second part of this enclosure calculates'the '
combination of NCS boron and xenon and samarium worths to ensure Mode 3 is maintained when the shutdown banks are out of the core. The required NCS boron to demonstrate this reactivity condition (as a function of T-AVG > 500'F - 557'F and core burnup) is obtained from the Data Book. The calcu-lated values of xenon and samarium worths are used (reduced by a conservative 10%) to supplement NCS boron to demonstrate adequate negative core reac-
-tivity to remain in Mode 3. The affect of this part of the enclosure is to show for what times Mode 3 will be maintained with shutdown banks withdrawn given current NCS boron and xenon / samarium behavior.
The third part of this enclosure calculates the combination of NCS boron and xenon worth which will correspond to core reactivity conditions which would '
result in criticality whit
- 750 pcm inserted (combination of RCCA insertion and samarium worth above equilibrium). This gives the operator guid-ance on the ~ times. dilution will be required to '
achieve a reasonable ECB/P and when borating the NCS to compensate for xenon decay (thus keeping the unit near the eventual ECB) is suggested. At all times shutdown margin will be maintained by the first part of the enclosure.
Enclosure 4.6: This enclosure calculated the boron conr.entration required to meet TS 3.9.1. The fuel cyc1s specific boron to maintain Keff < 0.95 (with 9._4
highest worth RCCA out, for RCCA latching) is
!_ compared with 2000 to see_which is higher. The
, higher value is then documented to be the value to
- j. be used as TS 3.9.1.
, One erroneous statement appears in the FSAR con-
- cerning this subject. -Section 4.3.1.5 states
- core h reactivity will remain < 1.0 if all.RCAAS were withdrawn during refueling. This condition is not verified by this enclosure _nor required by TS. f r
L Also, this jttgnificant procedure should ,be included ;
E on the FSAR list of OPS in Section 13.5.2.1.1. '
In all of the above enclosures, data are obtained i from the Data Book. These data are generated by~
Duke Nuclear Design by NRC - approved methods, i These data in combination with the above enclosures a maintain core reactivity within the assumptions of L
all applicable FSAR accidents. Thus the conse-quences of all FSAR accidents sensitive to core !
reactivity will be within that evaluated by the !
FSAR. Equipment related to safety is not affected. 3 MP/0/A/7150/66 Reactor Coolant Pump Hatch Covers Remeval and r Replacement: MP/ 0/A/7150/66, Reactor Coolant Pump Hatch Covers Removal and Replacement, provides a systematic and controlled means to gain access from !
the operating deck to the reactor coolant pumps in lower containment. The hatch covers and their gaskets-are a portion of the physical barrier that separates upper containment from lower containment, assuring LOCA energy is exhausted into the ice ,!
e condenser. This procedure is only for the removal ,
L and reinstallation of_ hatch covers provided in the ,
design of the containment building. Since the- .!
procedure makes no functional changes to.the hatch -*
covers or openings and since the appropriate noti-fication of station contacts is included, there are .
no unreviewed safety questions created and no Tech ;
Spec changes required. ;
PT/2/A/4150/21, Reissue Post Refueling Controlling Procedure for Startup-Testing: Significant changes included in reissue:
- 1) Added step and enclosure to calculate Hot Zero Power (HZP) differential boron worth; and Acceptance Criterion (115% of predicted) on ,
value obtained. Differential boron worth evaluation is recommended by ANSI /ANS-19.6.1-1985, Reload Startup Physics Tests for Pressurized Water Reactors. Data '
required for evaluation is obtained from PT/2/A/4150/10, Boron Endpoint Measurement, and
7 L
'[- PT/2/A/4150/11A, RCCA Bank Worth Measurement by Boration/ Dilution; therefore this evaluation requires no new testing. Duke Power-Design Engineering plans to revise Chapter 14 of FSAR to reflect standard reload physics testing for McGuire/ Catawba, whit.h will address evaluation of differential boron worth.
- 2) Added steps to have IAE perform Source Range .
Voltage plateaus if required at end of Zero l Power Physics Testing (ZPPT). Flux level is established above manufacturer's recommended minimum flux level of 1000 cps, which provides strong signal for obtaining plateau, while maintaining sufficient margin to Source Range 111 riux Trip Setpoint (10' cps). During work on one Source Range, Reactor Hi Flux protection '
) is provided by the other Source Range Channel, 2 Intermediate Range Channels, and 4 Power Range Channels (low and high power trips; high power trips set to 25% for ZPPT). Tech Spec 3.3.1 allows operation below P-6 with one Source Range out of service as long as opera-tions involving positive reactivity changes have been suspended (above P-6, source ranges are not required). Voltage plateaus will be obtained per approved IPs. Accident analyses in FSAR Chapter 15 are unaffected, t
- 3) Added Enclosure to install and remove reactiv-ity computer used during ZPPT. Requires removing one power range channel from service, which changes Power Range trip logic to 1/3 instead of 2/4. Removal- of one power range channel is aliowed per Tech Spec 3.3.1.
Reactor Hi Pux protection is provided by the other three power range channels (low power and high power trips; high power trips set to 25% '
for ZPPT). Channel is removed from and re-stored to service per approved procedure.
Accident analyses in FSAR are unaffected.
- 4) Added steps to have MES and Performance Test Group to perform post-mod testing on S/G 1evel tap modification. Testing will verify that the S/G 1evel control system operates as designed following the modification. Testing will be done using approved procedures with their own Safety Evaluations. The controlling procedure calls for the MES test to be performed at '8%
power (on CF Bypass Control Valves), ~40% (low end of CF Control Valves), and '80% (interme-diate on CF Control Valves). The Performance test will be performed at '40%, '80%, and 16
$100%. Recovery from Performance test, which requires a 10% step load decrease, will be done ;
using approved operating pror:edure and will be within recommended fuel maneuvering limits.
Accident analyses in FSAR are unaffected.
- 5) Added provision to allow power escalation to '
540% power before flux map taken between 20%
and 30% has been analyzed. This change poten-tially reduces power escalation time by 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The standard reload test program that Duke Power has committed to requires flux map before exceeding 40% power for the " low power" map between 20% and 30% and analyzing it before exceeding 40%. Margin to Tech Spec Fg and F willnotchangesignificantlybetween30%and 40%. Margin to limits will be very large at these low power levels. FSAR Safety Analyses '
are unaffected.
- 6) Changed enclosure to evaluate flux map results to include evaluation of detector errors and to correctly adjust indicated F Margin for maps taken below 50%. Evaluation0 0f detector errors is required per standard reload test program.
No Tech Specs or FSAR analyses are affected.
The margin to F n limit is used to determine limiting power Tevel. Margin is indicated in percent; normally the limit power level can be ,
calculated with P=0.5 instead of actual power level; therefore, if power is below 50%, 50%
should be used instead of power level of map to determine limiting power level. Evaluation of Fn will ensure that limits o' Tech Specs 374.2.2 will not be exceeded. No FSAR Analyses are affected.
1
- 7) Changed testing plateaus slightly (from '81% to
'78% and from '98% to '100%). Change from 81%
to 78% is to comply with_ standard reload test program, which requires an " intermediate" power map at $80% power. 81% had been specified in initial issue of procedure so base load opera- -
tion per Tech Spec 3/4.2.2 could be entered if required. This is no longer a consideration as base load W(Z) factors are not provided.
Plateau is still above 75% required for Incore/Excore Calibration (Tech Spec 3/4.3.1).
No FSAR-analyses are affected. Change from 98%
to 100% allows testing to be conducted at full power. Initial issue of procedure held at 98%
until NC Flow could be verified (through precision calorimetric) to be greater than Tech Spec limit before going to 100%. Any problem
t E
I with meeting Tech Spec 3/4.2.3 would be iden-F tified through low power and/or intermediate. ,
i power maps, and power would be limited accord-ingly. No FSAR analyses are affected.
- 8) Added steps while escalating from 50% to 78%
power to check capacity of one CF pump.
L, Operation with one CF pump is not outside of normal operating procedure up to 65%. However, these steps may require going above 65% with only one CF pump in header (the other pump will l'? be running but not providing flow). The second pump will be available if needed. Loss of feedwater flow accidents are analyzed in FSAR Chapter 15. Potential transients incurred while operating on one CF pump will be bounded by those analyses. No Technical Specifir.ations are affected. Procedure contains guidelines for aborting test based on indications avail-able in the Control Room. .These steps were performed on the Catawba simulator with no ,
unexpected responses.
- 9) Deleted step to perform Incore/Excore Calibra-tion during power escalation from 50% to 78% '
and added step to perform calibration at 78%. i Change was made because of unreliability of data obtained during power escalation in the past (Unit 2 Cycle 2, Unit 1 Cycles 2, 3, and 4). Change will ensure Tech Spec 3/4.3.1 i limits and core power distribution assumed in FSAR analyses will not be violated. Calibra- r tion is required by Tech Spec 3/4.3.1 to be performed above 75%, and recommended (by Westingouse) to be performed before exceeding 90% power.
-t
- 10) When checking calibration of excore detectors at 100%, changed criterion for requiring .
recalibration from 2% to 1% (using interim, or-I point, calibration). Also added step to do full calibration if error between Incore and Excore AFD exceeds 3%, as required by Tech Spec 3/4.3.1. Steps will ensure Tech Spec 3/4.3.1 limits and core power distribution assumed in FSAR analyses are not violated.
- 11) Changed instructions for establishing ZPPI test band so that a full decade is not required.
l This allows test band to be set higher in the case where nuclear heat is observed near low end (between 1.7 and 2.0) of decade, while maintaining a factor of 2 margin to the point of adding heat. Being able to set test band 1
98,
L1 L t
- . higher improves quality of flux signal from- l i- power range channel to reactivity. computer, and j
, thus enhances reactivity measurements. FSAR F Table 14.2.12-2, for initial startup testing, 4 specifies that upper limit will be approxi- .:
-mately 1 full decade below nuclear heat. This !
e section is being replaced with standard reload
, test program by Duke Design Engineering.- No !
Tech Specs or FSAR analyses are affected.
. The test program does appear significan't enough to' be included in FSAR; Duke Design Engineering is working on replacing FSAR Chapter 14 with Standard j Reload Test Program. This program has been submit- '
ted to the NRC. All changes have been verified to :
be consistent with this program.
The changes described above do not increase the .
probability or consequences of an accident described :
in FSAR, nor do they create the possibility of an unanalyzed accident. Plant equipment important to safety are not'affected. ,
PT/2/A/4400/01, ECCS Flow Balance: Part of this change allows the ;
Change #1 flow balance portion of the test to be performed t without testing pump performance by eliminating ,
steps to throttle flow to different test points. No.
accident scenarios are affected by this change since .
.ECCS is not required in no mode. No equipment t malfunctions are expected to occur or have their consequencess increased since all pumps will be operated within their design ranges. Also,' since ,
ECCS is not required during no mode, the margin of safety as defined in the basis of Tech Specs is not reduced.
Another part of change incorporates new flow balance criteria for the NV pumps cold leg injection. .This evaluation will not cover the implications of changing this criteria since Design Engineering has ;
already done this in their analysis. The change also adds a correction for NC system water level to the NI cold and hot legs and NV cold leg injection -li flows. This correction was provided by Westing-house. I&E will be doing MOVAT testing during the :
course of this' test. -At all times, the pumps will be provided adequate suction sources anJ adequate discharge flow paths. A smaller range gauge will be used to measure NI pump miniflow. This is a conservative change since greater accuracy will be realized. Higher range gauges will be used to measure cold leg flows but are acceptable per design analysis. ,
l
m ,
i, <
F- Since this test will not be performed with fuel in the core (no mode), no accident scenerios, either previously revi wed in the FSAR or not, are either increased or c eated. Since all pumps will be p operated with the required flow rates, no equipment i
malfunctions should occur. The margin of safety as defined in Tech Specs is not reduced by this change.
, OP/2/A/6700/01 Unit Two Data Book: OP/2/A/6700/01, Unit Two Data ic Change #90 Book, Table 2.2 contains Excore NIS Data for use by station personnel, The following describes the three components of the table and their use.
- FULL POWER CURRENTS:
h This section is used to record the Power Range NIS full power calit: aion currents at axial offsets of
+20%, 0% and -20% determined per approved procedure o PT/2/A/4600/05E, Refueling ENB Calibration. The M Factors denoted in this section have not been changed (they remain the same as the factors deter-f mined per the final Cycle 2-performance of the L quarterly incore/excore NIS Calibration). The new calibration currents are generated by adjusting the final calibration currents of Cycle 2 for the expected change in core power distribution from end of Cycle 2 to beginning of Cycle 3. These currents are used by IAE to enhance the calibration of the NIS to ensure the most accurate possible indication of AFD (and f,-(AI) input function to the OTAT-setpoints in the 7300 circuity) during initial power escalation to 30% F.P. following refueling (or higher if permissible by PT/2/A/4150/21, Controlling-Procedure for Startup Testing).
Y INTERMEDIATE RANGE NIS-TRIP SETPOINTS:
The trip setpoints for NIS Channels'N35 and N36 herein noted have been derived by PT/2/A/4000/05E in the same way the Power Range currents were deter-L mined. These sotpoints ensure conservative estab-lishment of the 25% Full Power Trip _setpoints-and ensures that they are at all times set at below the allowable limit of 31%, as set forth in Tech Spec Table 2.2-1.
POWER RANGE NIS TRIP SETPOINTS:
Trip setpoints for the Power Range Detectors may at p no time be deliberately established at greater than
- l. 109% Full Power. The trip setpoints may be set
? lower than 109% by approved procedure, Tech Spec Action Statement, or by direction of the Shift
? Supervisor. The value of 109% Full Power is set
-l IM
y: - 1 9
p l n
l forth by Tech Specs to ensure that power operation ;
is at all times bounded by the assumptions used in- .
4 FSAR Chapter 15 analyses. Any setpoints below 109% !
are for desired conservatism or compliance with Tech Specs. In this case the trip letpoints are estab-
- lished at 25% F.P. per Tech Spec.Special Test ;
Exception 3.10.3a approved procedure PT/2/A/4150/21, Post Refueling Controlling Procedure for Startup
- L Testing, for the purpose of imposing strict .
i conservatism on initial operation at below 10% F.P. '
following refueling. This ensures that all low l3 power (particularly zero power physics) testing is satisfactorily' completed,-thereby validating the new i- core's design predictions, before continued power ,
escalation is permitted, Data incorporated into Table 2.2 through this change I will not increase either the probability or conse-i quences of any accident or safety significant F equipment malfunction either analyzed or unanalyzed -
t by the FSAR. The margin of safety assumed in ,the L bases of all affected Tech Specs is not reduced in
, any way by this change.
TT/2/A/9200/52 NI Pump 2A Head Curve Test: FSAR Table 14.2.12-1 page 40 describes an ECCS test with the vessel head remcved. The-attached test performs a NI pump head test in Mode 5. The NI pump will take suction from ,
the FWST and inject throu the NI Cold Legs with ND in operation providing core cooling. The NC system ,
will be prevented.from going solid by dropping.
pressurizer level to 5% before initiating injection.
A vent path will be provided through two pressurizer '
e PORV's to the PRT and out through a PRT vent. The ,
vented gas will be routed to one of the Containment Ventilation Charcoal Air Filters to minimize air- -
borno contamination. The following considerations '
were made in this evaluation:
Reactivity concerns are satisfied since the source of water for this test is the FWST which-is maintained at >2000 ppm boron.,
LTOP concerns are satisfied by continuously .
venting the pressurizer to the PRT and the FWST liquid temperature being greater than 70 F which ensures Unit 2 NDT concerns are met. 'The test has ciiteria to stop the N1 pump when '
pressurizer cold cal level reached 90% to ensure the pressurizer does not go water solid.
Additionally, one PORV is adequate to maintain .
LTOP in the eye.1t of an NI pump starting with the NC system water solid.
101
'T C
V The PRT is protected by having a continuous
! vent path to a Containment Ventilation Charcoal Air Filter and by providing test termination criteria of PRT pressure - 60 psig. Addition-ally, the PRT is protected by a rupture disc which relieves at 100 psig.
The NI pump miniflow is verified to be adequate-
! by performing the IWp test prior to performance of this procedure as required by the procedures prerequisites.
Exceeding of runout for the NI pump is unlikely as backpressure is present from running the ND system for core cooling and static head from water level being at 5% in the pressurizer.
The system is balanced for 0 backpressure in the NC system. Additionally, backpressure will build through the test as the pressurizer is filled compressing the gas in the remaining pressurizer gas space and the PRT.
Once any of the above termination criteria are met, the pump will be stopped. Residual heat removal will not be affected and the minimum number of iniaction flowpaths will be maintained. FSAR Sections 6.3.2.2 (ECCS System) and Table 14.2.12-1 page 40 (ECCS System Tests) were consulted. The Chapter 15 Safety Analysis was also reviewed and is not affected.
TT/2/A/9200/16 CA System Autostart Transient Test: The purpose of this procedure is to determine the transient pres-sure characteristics in the Auxiliary Feedwater l System suction piping during a simultaneous auto-start i of CA Pumps 2A and 2B. .This data will be used by Design Engineering to determine if restrictions related to operating with the CA CST isolated may be lifted.
The CA System will be aligned for standby readiness with steam generator temperature and pressure at !
approximately no-load conditions. An automatic start signal will be initiated for CA Pumps 2A and 28 simultaneously. The resulting pressure transient i in the suction piping will be recorded for analysis.
Precautions-are taken to prevent introduction of RN water to the CA System. Valves 2CA15A, CA Pump 2A Suction from RN, and 2CA188, CA Pump 2B Suction from RN, will be maintained closed during the transient by placing their control switches to the " CLOSED" position. This will result, however, in both motor-driven CA pumps being technically inoperable for the duration of the test. This condition is allowed by-Tech Spec LCO 3.7.1.2, Action B. The 102
I h
turbine-driven CA Pump operability will be unaf-fected by this test. The motor-driven pumps may be made operable again by simply placing the control switches for 2CA15A and 2CA18B to the "AUT0" posi-tion. Both motor-driven CA Pumps are expected to be inoperable at the same time for no more'than five minutes at a time.
The probability of and consequences of an accident previously evaluated in the FSAR will not be in-creased. Only one CA Pump is required for any accident analyzed in the FSAR, and the turbine-
. driven pump will remain unaffected by this proce-dure. Although technically inoperable, the motor-driven CA Pumps are available and the suction valves
.to RN may be realigned at any time by the Control Room Operator. Since the CA System is in normal alignment.for this test, no new accident scenarios are created.
The probability of consequences of a malfu'nction of.
equipment important to safety will not be increased.
- The CA System will be aligned for standby readiness, .
which is the normal alignment for Modes 1, 2, and 3. i Should the turbine-driven pump malfunction, if required during the test, the motor-driven pumps would still be available and RN could be aligned ,
quickly by placing the switches for 2CA15A and 2CA18B to the "AUT0" position. '
The possibility of malfunctions of equipment impor- .
tant to safety different than those already evalu- ,
ated in the FSAR will not be created. The primt ~ ,
concern will be the period of time that the auto ,
matic swapover to RN for the motor-driven CA Pumps '
is isolated. The procedure states that should the loss-of-suction alarms stay in for five seconds, the. -
pumps are to be tripped and the automatic start
~
I logic reset immediately, ,
The margin of safety as defined in the Tech Specs is '
not reduced. The minimum required number of CA l Pumps will be operable and available at all times, u .'
PT/2/A/4600/05E Refueling ENB Calibration: PT/2/A/4600/05E calcu-lates preliminary Power Range NIS calibration .
currents and Intermediate Range NIS Rod Stop and j j Reactor Trip setpoints, prior to startup following ,
refueling, based upon the core power distribution i measured near the end of the previous fuel cycle and the core power distribution predicted for the i beginnning of the new fuel cycle. These calibration l currents and setpoints are derived by adjustment of i . existing calibration data (from the previous cycle)
- with a ratio of the measured and predicted core power distribution data,
! J03
7 C
b The NI pump miniflow is verified to be adequate 1 by performing the IWP test prior to performance i
of this procedure as required by the procedures prerequisites.
Exceeding of runout for the NI pump is unlikely as backpressure is present from running the ND system for core cooling and static head from water level being at 5% in the pressurizer.
The system is balanced for 0 backpressure in the NC system. Additionally, backpressure will build through the test as the pressurizer is e filled compressing the gas in the remaining pressurizer gas space and the PRT.
Once any of the above termination criteria are met, h the pump will be stopped. Residual heat removal will not be affected and the minimum number of injection flowpaths will be maintained. . FSAR Sections 6.3.2.2 (ECCS System) and Table 14.2.12-1 page 40 (ECCS System Tests) were consulted. The Chapter 15 Safety Analysis was also reviewed and is not affected.
TT/2/A/9200/16 CA System Autostart Transient Test: The purpose of this procedure is to determine the transient pres-sure characteristics in the Auxiliary Feedwater System suction piping during a simultaneous auto-start of CA Pumps 2A and 28. This data will be used by Design Engineering to determine if restrictions related to operating with the CA CST isolated may be lifted. I The CA System will be aligned for standby readiness with steam _ generator temperature and pressure at approximately no-load conditions. An automatic start signal will be initiated for CA-Pumps 2A and.
2B simultaneously. The resulting pressure transient in the suction piping will be recorded for analysis.
Precautions are taken to prevent introduction of RN water to the CA System.- Valves 2CA15A,.CA Pump ~2A Suction from RN, and 2CA18B, CA Pump 28 Suction from RN, will be maintained closed during the transient by placing their control switches to the " CLOSED" position. This will result,. however, in both motor-driven CA pumps being technically inoperable ,
for the duration of the test. This condition is j allowed by Tech Spec LCO 3.7.1.2, Action B. The turbine-driven CA Pump operability will be unaf-fected by this test. The motor-driven pumps may be made operable again by simply placing the control i
102
I f-switches for 2CA15A and 2CAIBB to the "AUT0" posi-tion. Both motor-driven CA Pumps are expected to be inoperable at the same time for no more than five F minutes at a time.
The probability of and consequences of an accident L previously evaluated in the FSAR will not be in-creased. Only one CA Pump is required for any accident analyzed in the FSAR, and the turbine-driven pump will remain unaffected by this proce-dure. Although technically inoperable, the motor-driven CA pumps are available and the suction valves to RN may be realigned at any time by the Control Room Operator. Since the CA System is in normal alignment for this test, no new accident scenarios
, are created.
The probability of consequences (if a malfunction of equipment important to safety will not be increased, r The CA System will be aligned for standby readiness, which is the normal alignment for Modes 1, 2, and 3.
Should the turbine-driven pump malfunction, if required during the test, the motor-driven pumps would still be available and RN could be aligned quickly by placing the switches for 2CA15A and 2CA180 to the "AUT0" position.
The possibility of malfunctions of equipment impor-tant to safety different than those already evalu-t' ated in the FSAR will not be created. The primary concern will be the period of time that the auto-matic swapover to RN for the motor-driven CA Pumps is isolated. The procedure states that should the -
loss-of-suction alarms stay in for five seconds, the !
pumps are to be tripped and the automatic start logic reset immediately.
The margin of safety as defined in the Tech Specs is not reduced. The minimum required number of CA Pumps will be operable and available at all times.
PT/2/A/4600/05E Refueling ENB Calibration: PT/2/A/4600/05E calcu-lates preliminary Power Range NIS calibration currents and Intermediate Range NIS Rod Stop and Reactor Trip setpoints, prior to startup following refueling, based upon the core power distribution measured near the end of the previous fuel cycle and the core power distribution predicted for the beginnning of the new fuel cycle. These calibration currents and setpoints are derived by adjustment of existing calibration data (from the previous cycle) with a ratio of the measured and predicted core j power distribution data.
1
I
- There are two significant changes resulting from the reissue of this procedure. The first involves the calculation of predicted Power Range NIS calibration currents at Core Axial Offsets of +20% and -20%.
This additional calibration data (the 0% Axial Offset calibration currents are already calculated per this procedure) is required to be documented in the affected Data Book Table and enhances the preliminary Power Range NIS calibration performed prior to unit restart following refueling. 'It also
! allows IAE to perform a thorough checkout of the Control Room al Meters. Unit restart with the best i
achievable Power Range NIS calibration allows power escalation up to 30% F.P. with confidence.in indi-cated Axial Flux Distribution (AFD). This is i critical to compliance with the Tech Spec limits on the F and F core peaking factors, as well as AFD, QPTR,0andin$catedpower. It also enhances the validity of inputs.to the f3 (AI) functions of the OTAT setpoints (described in Tech Spec Table 2.2-1) which serves important safety functions in a variety of FSAR Chapter 15 analyses.
The second change allows I and E to waive the adjustment of the I/R NIS Rod Stop and Reactor Trip setpoints if the calibration currents calculated per this procedure correspond to setpoints which are within 3% (of full power), on the high (conserva-tive) side of the present setpoints. This is permissible with regards to the Nuclear Safety functions of these setpoints since they would be impacted at a lower reactor power level. This conservatism ensures that no safety analyses are affected by this change. Adjustment of these setpoints would be performed in all cases in which the new calibration currents are lower (and there-fore nonconservative) than the present 20% F.P. and 25% F.P. setpoints. Allowing the actual I/R NIS Reactor Trip setpoint to be inaccurate up to 3% (of full power) in the conservative direction (i.e. 22% 1 F.P.) will not result in an unanticipated Reactor Trip when the I/R trip function is enabled by P-10 j (Power Range NIS 5 10% F.P.) during power reduc- -
tions. This assessment is based upon an evaluation of the expected changes in I/R and P/R NIS response due to heavy control rod insertion during power reductions. This evaluation concludes that it is not possible for the indicated power by the I/R NIS to be 2 22% F.P. while the P/R NIS is indicating 5 10% F.P.
Neither the probability nor the consequences of any accident or safety significant equipment malfunc-tion, either previously analyzed or unanalyzed by '
the FSAR, will be increased by this change. This 4
i I
r evaluation is based on the fact that conservative actuation of the I/R NIS Rod Stop and Reactor Trip setpoint safety functions are not degraded by this change. Likewise, the OP6T and OTAT trip functions are not degraded, but rather, enhanced by the improved accuracy of preliminary P/R NIS calibra-tion. The margin of safety defined in the applica-
, ble Tech Spec bases are not reduced in any measure by this change. CompliJnce with Tech Specs ad-dressing core power distribution is enhanced by the more thorough P/R NIS preliminary calibration.
PT/2/A/4600/068 Incore Detector Setpoints Determination: This change is restricted to Unit 2 Cycle 3 startup for ,
the purpose of handling the special situation posed l i by the shortening of ten incore flux thimbles by '
repositioning during the refueling outage. Vibra-tion induced through wall wear measured by eddy current testing required that these thimbles be shortened from 2 inches to 3.25 inches (depending on the measured wear scar geometry). Normally this procedure establishes top of core setpoints at 2 inches below the actual end of the flux thimbles, however, this practice with shortened thimbles could result in the failure of the ENA detectors to completely survey the entire height of active core ,
(the uppermost of the 61 axial locations surveyed I could be missed). This change allows the top of l core setpoints to be established at 0.5 inches below !
the actual end of flux-thimbles to ensure that the i active core-height is adequately surveyed. j Neither the probability nor the consequences of any accident or safety significant equipment malfunc-tion, either analyzed or unanalyzed by the FSAR, will be increased by this change. The only plant equipment directly impacted are the flux thimbles and ENA detectors. Establishment of top of core setpoints 0.5 inches below the ned of the thimbles i will still ensure protection of the detectors from l butting the ends of the thimbles during normal operation. The core peaking factors (F and F3g),
which are meoured by the incore detect rs per periodic surveillance to verify compliance with Tech !
Specs and thereby validate the core power distribu- 2 tion assumptions used by Chapter 15 accident analy-ses in the FSAR will continue to be accurately monitored following implementation of this change.
Likewise, the axial power distribution will continue ;
to be accurately measured for the purpose of main- x taining a valid calibration of the Excore NIS (for the purpose of monitoring core Axial Flux Differ-ence). This ensures that compliance with the Tech Spec RAOC limits is verified, thereby validating the i associated assumptions made in the FSAR accident I
l LOS ,
analyses. Any degradation of the ENA System's .
capability to monitor the entire core height would I be restricted to axial location 61 only (at the top of the core). Loss of incore flux data at this elevation would have no impact on the incore com-puter code's (SND Core Program) ability to calculate core parameters due to the fact that raw incore data from axial location - 61 is usually invalid due to a neutron reflection peak at the top of the core, f This data is therefore routinely smoothed by the incore code. In the absence of data from location 61, the SND Core program would extrapolate a value for it, as is normally done. The margin of safety '
as defined in the bases of any applicable Tech Specs will not be decreased in any way by the determina-tion of detector setpoints under this char.ge. This is due to the fact that the capabilities of the Incore Flux Mapping System are not degraded in any measure by it.
PT/2/A/4250/03C, Turbine Driven Auxiliary Feedwater Pump #2 Change _#27 Performance: The purpose of this change is to ensure that CA Pump #2 is sufficiently challenged by a cold, quick start on a quarterly basis. This change is in response to recommendations made in the ,
Operational Readiness Review conducted by INPO in January of 1989. This change will result in the pump not being manually started up in minimum speed with the excess moisture drains open. Instead, the pump will start up at full speed just as it would on :
an emergency start signal.
The basic test method and expected results are not being affected as a result of this change. The rest of the system is not affected since the pump is aligned in recirculation and is isolated from the safety related portions. The step to ensure the alignment per the Operations procedure is not necessary since a correct valve lineup is provided on Enclosure 13.5. Therefore, the probability of and consequences of an accident previously evaluated in the FSAR are not increased. In addition,'no new accident scenarios are created since the valve alignment and test method are not revised as a result of this change.
The pump will be started as it is designed to be on an emergency signal. It will also be run well within the design flowrates and speeds. Therefore, the probability of and consequences of a malfunction of equipment important to safety are not increased. ,
Since no change to pump operating parameters is made, no new possibilities of equipment malfunctions are expected to occur.
10 6
t Since no change is made to the procedure as to how many operable CA pumps are in service, the margin of safety as defined in the bases of the i Technical Specifications is not reduced.
PT/0/A/4450/010, Control Room Area Outside Air Pressure Filter Trains Change #8 Performance Tests: This change reduces the accep- <
tance criteria for the bypass leakage of the HEPA and Carbon Adsorber Filter banks on ICRA-PFT-1 to the values used on 2CRA-PFT-1. These filter units service the same area and are technically A and B train so the acceptance criteria should be the same.
The changed value allows less leakage in ICRA-PFT-1, ,
therefore it is more conservative than the Tech Spec value so no change to Technical Specifications is ,
required. The consequences of an accident are not increased, the probability of an accident is not increased and no unreviewed safety question is created.
PT/0/A/4450/17, Safety Related Filter System Run Time Monitoring:
Change #9 This procedure change allows the testing of carbon samples on VE to return to its normal surveillance per Tech Spec 4.6.1.8. The carbon testing was done every 60 days (per DE recommendation) because the heater wiring was not qualified for Post-Accident radiation. The heater wiring has been replaced under exempt changes CE-1502 and CE-1503 so the 60 day testing is no longer required..
Reference:
letter R. R. Weidler to J. W. Hampton, 6/11/87, CN-1211.00-21 and letter F. N. Mack to W. F. Beaver, 5/19/89, CN-175.50 and CN-217.45.
This change also reduces the acceptance criteria for VC carbon lab analysis to 0.175% penetration as required by Regulatory Guide 1.52, Revision 2, March 1978 paragraph C-6-a - This value allows less leakage through the carbon, therefore it is more conservative than the Tech Spec value (of 1%) so no change to Techaical Specifications is required. For this reason ar.d the reasons stated above, the conse-quences of an accident are not increased, the probability of an accident is not increased and no unreviewed safety question is cretted by these changes.
~
PT/2/A/4600/05C Post Refueling Incore/Excore Calibration: The purpose of PT/2/A/4600/05C, Post Refueling Incore/Excore Calibration, is to collect incore flux map data and excore power range data at various Axial Offsets during the initial power escalation following a refueling outage and to determine calibration currents and gain (MJ factor) for each I
of the excore power range channels in order to make excore axial flux difference (AFO) consistent with 1 01 L
incore measurements. Another result of the cali-bration is to normalize excore tilt ratios to e approximately 1.000.
Excore AFD and tilt ratios are the primary real time indication of Core Power Distribution. There are ;
Technical Specifications associated with both AFD ,
and Quadrant Power Tilt Ratios (QPTR), which is' -
defined as the maximum excore tilt ratio. Operation within the Technical Specification Limit on AFD and QPTR ensures that the core power distribution between the monthly flux maps remains within the power distribution assumed in the FSAR Accident Analyses. A function of excore AFD, f(delta I), is included as a term in the over temperature delta temperature (OTDT) trip setpoint formulation. The f(delta I) function reduces the OTDT trip setpoint for axial power distributions that are more positive ,
or more negative than a specified dead band.
The procedure is written to ensure that the required f incore and excore data is obtained safely and efficently. The AFD oscillation is intentionally induced to allow measurements over a wide range of axial power distribution in order to determine a relationship between incore and excore measurements that will remain valid at extreme axial power distributions. The negative test limit for AFD is established with a minimum margin to Technical Specification limit of 2% to ensure that the assumed power distribution in FSAR analyses is limiting..
A number of steps are taken to ensure that calibration data is accurate. Incore and excore data are analyzed using approved and controlled software. A least squares fit of the incore/excore relationship is performed for each power range detector, and a fit correlation and error for each point used in the fit are calculated so that bad points may be detected and excluded from the fit.
Calibration data is used to calculate excore AFD using data from the first flux map and compared to incore results to ensure consistency. Tilt ratios l are also calculated in like manner and verified to be approximately 1.000. After IAE has performed the calibration of the power ranges and calibration data required for the OAC Excore Power Distribution l monitor have been entered, power range currents are obtained concurrently with 0AC data. AFD and quadrant tilt ratios from manual calculations are l compared to the OAC indications to verify consis-l tency.
L I
The test does not require any unanalyzed operation with respect to core power distribution. Systems 108 l
i p
used to support test, such as the rod control system i and the chemical and volume control systems, are used in normal operational modes per approved operating procedures. The movable incore detector ,
system is also used in accordance with an approved operating procedure. Excore detector data is obtained without any abnormal configuration of power range channels, they remain fully functional throughout data gathering portion of test. Obtain- !
ing excore current data from a power Range channel has sometimes (very infrequently) resulted in ,
receiving a Rate Trip on that channel. This has no effect on the Unit unless another channel is tripped, in which case a Reactor Trip will occur. A limit and precaution is included to remind the user of this possibility and to specify that the user should verify that all channels are in Normal Status (i.e. not tripped) before opening a power range drawer. If a Reactor Trip were to occur, the transient would be bounded by analyzed transients.
Calibrations of the power range channels are done in accordance with approved instrumentation procedures which require that the channel being calibrated be ,
removed from service and asociated bistables (
tripped. Reactor trip logic is thereby reduced to i 1/3 logic, which is conservative with respect to the normal 2/4 logic.
For the above reasons, it is concluded that the probability or consequence of an accident evaluated in the FSAR will not be increased, nor will the possibility of an accident not evaluated in the FSAR be increased. Likewise, the probability or conse-quences of a malfunction of equipment important to safety evaluated in the FSAR will not be increased, ,
and the possibilty of an unevaluated malfunction will not be increased.
As operation will remain within Technical Specif1-cation limits at all times during the test, the !
neargin of safety will not be reduced.
PT/2/A/4150/11B Control Rod Worth Measurement By Rod Swap: The only change incorporated by this reissue with potential ,
Nuclear Safety Significance is the addition of guidelines for the performance of rod swaps starting with the Reference Bank fully inserted and the rod bank of lowest worth slightly inserted. This i initial condition can result from boron dilution overshoot during the integral rod worth measurement '
of the Reference Bank. Normally, rod swaps are conducted from an initial condition of Bank RF <40 pcm of the fully inserted position and all other rod 109
7 Q I:
9 , w n ,
e banks fully withdrawn. In this case, the critical position of Bank RF (with all other. banks withdrawn) .
is noted prior to, and following sequential exchange. !
? of the seven banks with Bank RF-in order to-infer i their reactivity worths. The initial and final critical positions of Bank RF are then evaluated and the change (i.e. reactivity drift) is equally applied over the inferred worths of the seven banks by_either incrementally adding (in event of inad-vertent dilution of the NCS) or deducting (in event '
of inadvertent boration of the NCS) reactivity from the inserted integral worth of Bank RF (with all other banks withdrawn) before each of_the seven rod n ,
swaps. ,
In order to handle the new situation, the inserted worth of the lowest worth bank (Bank 1) is measured prior to commencing rod exchange. Following com- '
pletion of the: exchanges, the reactivity difference a (drift) between the final Bank I critical position and the initial critical position is measured. This reactivity worth is then equally applied, in con-junction with the initial inserted worth or Bank 1,
.over the seven banks measured by rod exchange to adjust for the drift noted over the-test interval.
This adjustment involves addition of the reactivity contributions of the insertion of Bank 1 to the Total Measured Integral Worth of the Reference Bank.
.t Addition of these r.aw guidelines will save up to two hours of critical path time, which would have been expended by borating Bank 1 out before initiation of rod exchanges. These guidelines ensure that accu-racy of inferred rods worths by rod exchange is enhanced ~by drift adjustment. This is important to the accurate validation of predicted core design '
parameters.
Neither the probability nor_the consequences of'any accident or safety significant equipment malfunc-tion, either analyzed or unanalyzed in the FSAR will be increased by the changes incorporated by this ,
reissue. Adequate Shutdown Margin is assured (as required by Tech Spec 3.1.1.1) with the additional. 3 insertion of Bank 1 per the preliminary evaluation performed by Enclosure 13.2 of PT/2/A/4150/21, Post Refueling Controlling. Procedure for Startup Testing.
This Enclosure deducts 40 pcm (in addition to the Total Worth of the Reference Bank, plus 15% percent to account for uncertainty) from the total available.
rod worth (minus 10% for uncertainty) to verify that available Shutdown Margin >1300 pcm (1.3% K/K) is available during rod swap testing. The margin of 110
r-"- ,
safety as defined in the bases of affected Tech Specs is therefore not reduced by this change.
MP/0/A/7600/104 Walworth Pressure Seal Globe Non-Return Valve ,
Corrective Maintenance: This procedure performs-corrective maintenance on Walworth non-return type pressure seal globe valves. Instruction manual CNM- ,
1205.00-0366 provided technical information in the development of the procedure. Plant specific information that applies to Catawba Nuclear Station Lwas also included in its development. .,
This evaluation is for changes made during the procedure upgrade process. These changes are in the procedure format and are not significant in content, The Catawba FSAR and Technical Specifications have been reviewed and are not affected by this proce- '
dure. This procedure applies to valves used in the r steam supply line to the aux, feedwater pump tur-
- bine-. The procedure will be used to correct and improve the performance of the valve and will maintain the~ valve within its original design requirements and specifications. The safety stan-dards previously specified within the FSAR and Technical Specifications will be maintained. The probability of an eccident or malfunction previously '
addressea in the FSAR will not be increased.' No unreviewed safety questions are involved.
MP/0/A/7600/91 Valcor Check Valve Corrective Maintenance: This procedure performs corrective maintenance on Valcor Check Valves. Instruction manual CNM 1205.08-0022 provided technical information in the development of
~
the procedure. Plant specific information that applies to Catawba Nuclear Station was also included in its development.
This evaluation is for changes made during the procedure upgrade process. These changes are in the procedure format and are not significant in content.
The Catawba FSAR and Technical Specifications have been reviewed and are not affected by this proce-dure. This procedure applies to valves used in the '
compressed air and waste gas system at Catawba.
This procedure will be used to correct and improve the performance of the valve and will maintain the valve within its original-design requirements and specifications. The safety standards previously specified within the FSAR and Technical Specifica-tions will be maintained. The probability of an accident or malfunction previously addressed in the FSAR will not be increased. No unreviewed safety questions are involved.
111
3
'MP/0/A/7600/35 Pacific-3" and 4" (#150) Globe Valve Corrective Maintenance: This procedure performs corrective j maintenance on Pacific 3" and 4", 150 pound, globe-
- valves. Instruction manual CNM-1205.00-0318 pro-vided technical information in the development of the procedure. Plant specific'information that applies to Catawba Nuclear Station was also included o in its development.
This evaluation is for changes made during the procedure upgrade process. These changes are in the procedure format and are not significant in content.
-The Catawba FSAR and Technical Specifications have been reviewed and are not affected by this proce-dure. This procedure applies to valves used in various plant applications. This procedure will-be used to correct and improve the performance of the valve and will maintain the valve within its original design requirements and specifications.
The. safety standards previously specified within the -
FSAR and Technical Specifications will be -
maintained. The probability of an accident or malfunction previously addressed in the FSAR will not be increased. No unreviewed safety questions are involved.
MP/0/A/7600/28B Borg Warner Pressure Seal Motor Operated Gate Valve Corrective Maintenance: This procedure performs corrective maintenance on Borg-Warner pressure seal gate valves with motor operators._ Instruction j manual CNM 1205.00-0311, -0323, and -1102 provided technical information in the development of the procedure. Plant specific information that applies
-to Catawba Nuclear Station wa3 also included in its development.
This evaluation is for changes made during the procedure upgrade process. These changes are in the' :
procedure format and are not significant in content. '
The Catawba FSAR and Technical Specifications have 3 been reviewed and are not affected by this_proce- '
dure. This procedure applies to' valves used in _
various plant applications. This procedure-will be used to correct and improve the performance of the valve and will maintain.the valve within its original design requirements and specifications.
The safety standards previously specified within the g FSAR and Technical Specifications will be maintained. The probability of an accident or ;
malfunction previously addressed in the FSAR will not be increased. No unreviewed safety questions are involved. !
1 PT/2/A/4350/02E, CA, CF, and Turbine Interlocks Periodic Test: This Change #23 procedure change is being written to allow retesting 12
W '
w ,
G 4
of the~1ogic to valves 2SA2.and 2SAS without actu-ally running the CA Turbine-Driven Pump.- Steps requiring starting and response-time testing of the pump are deleted, and block valves 2SA1 and 2SA4 are added to the valve lineup and are verified closed.
Since steam to the CAPT will be isolated during the test, the CAPT will be inoperable. However, the- .
10CFR50.59 analysis for the procedure (dated 5/1/89) '
has already addressed CA pump operability (since the procedure normally aligns the CAPT to-run in recir-culation, which also makes the pump inoperable).
Therefore, the probability / consequences of any new or previously-analyzed accident or equipment mal- i function is not increased, and the m'argin of safety i is not reduced.
PT/2/A/4350/02E, CA, CF, and Turbine Interlocks Periodic Test: This ;
Change #24- procedure change is being written to allow retesting l of the logic to valves 2SA2 and 2SA5 without actu- !
ally running the CA Turbine-Driven Pump. Steps requiring starting and response-time testing of the ,
pump are deleted, and block valves 2SA1 and 2SA4 are added to the valve lineup and are verified closed.
Since steam to the CAPT will be isolated during this
- test, the CAPT will.be inoperable. However, the .l 10CFR50.59 analysis for the procedure (dated 5/1/89) l has already addressed CA pump operability (since the procedure normally aligns the CAPT to run in recir-culation, which also makes the pump inoperable).
Therefore, the probability / consequences of any new or previously-analyzed accident or equipment mal-function is not increased, and the margin of' safety.
is not reduced. ,
PT/1/A/4350/03C,- Turbine-Driven Auxiliary Feedwater Pump #1 Perform-Change #40 ance Test: This change has two parts. The first part is to add a column for locked valve verifica-tion. The.second part is to add ICA19, CA Pump #1 Normal Suction Isolation Valve, to the valve . lineup.
Part 1 is in response to violation 413, 414/87-30-03. ,
This is only an editorial change to ensure that Operations verifies that normally locked valves are L . returned to the locked position. This' portion does 3 ,
not affect the performance or outcome of the test and-therefore does not require any further analysis.
Part 2 is to add an additional valve to the lineup for CA Pump #1. The valve is the manual suction isolation to pump and should be in the lineup as a p 113 s.U a r l ,
w- 1 good practice. The addition of this valve an'd its 1 correct position does not have any affect on previ-ously' considered accidents or on any non postulated ones. -This~ change will actually decrease the change of an equipment failure since it further ensures that the pump will have an adequate suction path. ,
CA Pump #1 is inoperable during this test due to the system being aligned to recirculation. This change does not cause any further inoperability of .the CA System and therefore does not reduce the margin of safety as defined in the bases of the Technical.
Specifications.
PT/1/A/4350/03B, Auxiliary Feedwater Motor Driven Pump 1B Perform-Change #20- ance Test: This change has two parts. The first-
.part is to add a column for locked valve verifica-tion. The second part is to add ICA30, CA Pump 1B Normal Suction Isolation Valve, to the valve lineup.
.Part 1 is in response to violation 413, 414/87-30-03. This is only an editorial change to ensure that Operations verifies that normally locked valves are returned to the locked position. This-portion does not affect the performance or outcome of the test and therefore does not require any further analysis.
Part 2 is to add an additional valve to the lineup :i for CA Pump 18. The valve is the manual suction 1 isolation to pump _and should be in the 11eup as.a good practice. The addition of this valve and its correct position does not-have any affect on.previ-ously considered accidents or on any non postulated ones.- This change will actually decrease the chance of an equipment failure since it further ensures that the pump will have an' adequate suction path.
CA Pump IB is inoperable during this test due to the system being. aligned to recirculation. This change -i does not cause any further inoperability of the CA '
System and therefore does not reduce the margin of
-safety as defined in the bases of '.he Technical 1 Specifications, pT/2/A/4200/01R, M301, M234, M452 Penetration Leak Rate Test: The Change #6 purpose of procedure PT/2/A/4200/01R (M301, M234, i M452 Leak Rate Test) is to measure the leakage rate- f through the spare penetrations listed above by '
pressurizing with air between the two blind flanges of each penetration and using the air flow makeup method. Each spare penetration consists of a containment sleeve with a blind flange (with gasket)
IM
t ,
y-bolted to a welded flange at each end of the con- l tainment sleeve.
ANSI N45.4-1972 specifies that for major repairs to the containment structure. involving strength weld-ing, a pressure test shall precede leakage rate testing. Change #6 to PT/2/A/4200/01R provides for a structural integrity test on penetrations M371 and
'M394, both of which are having flanges rewelded.
j The structural integrity test involves maintaining a
. pressure of 17.0 psig.(vs. test pressure of 15.4' l psig) for 10 minutes.
The containment sleeve portion of the spare pene-trations has an ultimate internal pressure (1'e. the maximum pressure prior to failure) of greater than 1000 psig (FSAR Table 3.8.2-9). Any degradat'on in the flanges (or in the bolts which secure the-flanges) caused by the structural integrity test pressure will be apparent during the subsequent leak rate test via an increase in measured leakage rate, as. indicated by the Volumetrics Leak Rate Monitor.
For these reasons, increasing the test pressure by approximately 1.5 psig for 10 minutes will not increase the probability / consequences of any'acci- ;
dent or malfunction of equipment, nor will it create the possibility of any new accident or malfunction of equipment not previously evaluated in the FSAR, and the margin of safety will not be reduced.
PT/2/A/4200/01P, NF. Penetration Leak Rate Test: The purpose of Change #4 procedure PT/2/A/4200/01P (NF Penetration Leak Rate Test) is to measure the leakage rate through the "
spare penetrations listed above by pressurizing with ,
air between the two blind-flanges of each penetra- '
tion and~using the air ~ flow makeup method. Each spare' penetration consists of'a containment sleeve with a blind flange (with gasket) bolted-to n welded flange at each end of the containment sleeve.
ANSI N45.4-1972 specifies.that for major repairs to the containment structure involving strength wele ing, a pressure test shall precede leakage-rate testing. Change #4 to.PT/2/A/4200/01P provides for a structural integrity test on penetrations M371 and M394, both of which are having flanges rewelded.
The structural integrity test involves maintaining a r pressure of 17.0 psig (vs. test pressure of 15.4 psig) for 10 minutes.
The containment sleeve portion of the spare pene- ,
trations has an ultimate internal pressure (i.e. the '
maximum pressure prior to failure) of greater than 1000 psig (FSAR Table 3.8.2-9). Any degradation in 115
' +
q l
the flanges (or in the bolts which secure the flanges) caused by the structural integrity test pressure will be apparent during the subsequent: leak rate test via-an increase in= measured leakage rate, as indicated by the Volumetrics' Leak Rate Monitor.
For these reasons, increasing the test pressure by approximately 1.5 psig for.10 minutes will not '!
increase the probability / consequences of any acci-dent or malfunction of equipment, nor wi.11 it create the possibility of any.new accident or malfunction-of equipment not previously evaluated in the FSAR, and the margin of safety will not be reduced.
i PT/1/A/4200/31 SV Valve Inservice Test-(QU): This procedure, PT/1/A/4200/31 SV Valve Inservice Test - Quarterly, !
is a Procedure Writer's Guide upgrade of an existing procedure. This procedure is used to satisfy Technical Specification 4.0.5 stroke time require-ment in the time specified by Tech Spec Table 3.6-2a. The valves stroke timed are ISV-1, ISV-7, ISV-13,'and ISV-19 which are main steam (SM) power ;
operated relief valves (PORV). These valves are i designed to: '
I
- 1) Prevent lifting.of the Code Safeties on each steam line during mild transients;
- 2) Assist in reseating the safeties;
- 3) Provide a means for cooldown when the steam L dump system is unavailable; ,
- 4) Provide a safety grade means of cooldown to ND initiation; ;
l J
- 5) Close on any main steam isolation signal.
l This test is written to be performed 'in any mode. l The test allows any valve to be tested individually
- ' or all 4 to be tested at once. The valve (s) to be tested has its associated block valve closed, is opened, then closed by the closure of a switched j jumper installed across the SSPS output relay contacts for main steam isolation. From the time
'this switch is closed until it is removed and the PORV's are reset, NONE of the PORV's will be capable of opening automatically. They will ALL be capable of opening in manual modo at arv time during the test. Therefore, the safety functie n 4 item 4 above is not' diminished . The ability to provide main steam isolation'at any time during the test is not diminished. For any valve (s) testtd the closed block valve will provide isolation durfeg the time '
the PORV is open; It the isolated POM[s) is needed 116
y ,; -
m >
i i
t for cooldown to ND' initiation, the block valve can' ;
be opened and cooldown initiated.
The probability and consequences of an accident or a malfunction of equipment important to safety previ- j ously evaluated in the FSAR will not be increased.
The possibility of an accident or malfunction of equipment important:to safety not previously evalu-ated in the FSAR will not be created. The margin of a safety in the bases to. Tech Specs will not be reduced.
OP/2/A/0150/08 Rod Control, Change #5: This change implements the !
revision described in exempt change VN CE-1964, See the 50.59 evaluation for this exempt change.
MP/0/A/7150/48 Steam Generator Power Operated Relief Valve Correc-tive Maintenance: The purpose of this procedure change is to incorporate the modifications to be made to the S/G PORV's through the implementation of-NSM #CN-50395, Rev. O and 1. Potential. problems with the S/G PORV's requires these modifications for continued reliable operation. These modifications basically consist of 1)' Increased Pilot Port capacity and 2) An improved two piece piston ring design. Maintenance performed IAW MP/0/A/7150/48 , ,
will ensure continued reliable operation of the S/G '
PORV's.
The S/G PORV's are provided in the safety related portion of each main steam line upstream of the MSIV. These PORV's 1) Prevent lifting the SV-Code Safety Relief Valves during mild pressure tran-sients, 2) Assist actuated Code Safeties in reseat-ing, 3) Provide a means for plant cooldown when the
' steam dump system is unavailable, 4) Provide a safety-related means of achieving Reactor Coolant System Cooldown, 5) Help mitigate the consequences of a S/G Tube Rupture accident, and 6) Provide containment isolation when required.
The relieving characteristics and functional opera-tion of the valve will not be adversely affected.
Since the relieving characteristics will not be changed, and the operation of the valve to perform ,
its function will be enhanced, the probability-or '
consequences of a malfunction of equipment important to safety, or an accident previously evaluated in the FSAR will not be increased. No new equipment is affected, therefore, the possibility of malfunction of equipment irrportant to safety different than any already evaluated in the FSAR is not created.
Since no plant parameters or setpoints are altered by this procedure, the margin of safety as defined 13Z
y -
f t
i t.
in the bases to any of the Technical Specification's is not-reduced. There are no unreviewed safety i questions associated with this procedure revision. -[
. IP/0/B/3250/10, DRPI Alignment and Functional Test: This change is Change #6 intended as a precautionary. measure to ensure that NC Boron concentration is sufficient to keep the Reactor in Mode 5 (Kef f < .99) during the DRPI i Alignment Test. Per Tech Specs, only one bank of rods can be withdrawn at a time. This change will 1 in no way affect the probability or consequences of an accident as described in the FSAR. The Rod -
l
- 1 Control System and DRPI are not affected by this change. Also, no other equipment important to s, safety are affected. The margin of. safety as i defined in Tech Specs will not be reduced by this j change. Tech Specs requires Keff < .99 per Table 1.2; -this change will ensure this limit is not exceeded.
OP/2/A/6250/01? . Restricted Procedure Change Condensate and Feedwater System: This restricted change is being made to i provide guidance for full D/P stroke verification of. .,
valves 2CA-149, 150, 151, and 152. These verifica- 1 tions will be done in Mode 5 and will not affect l reactor safety. This change will not affect any j accidents previously evaluated in the FSAR, nor will the possibility of an accident.different than any evaluated in the FSAR be created. .
Equipment safety will not. be affected. CF pump speed will still be protected by an overspeed trip.
CF pump miniflow will still exist with 2CF-13 (CF Pump 2B Recirc Ctrl) in Auto. CF pump runout is not expected. S/G overpressurization is not expected l because vacuum.wi-11 be lined up on each S/G being tested. S/G overfilling should not occur because-procedurally the Feed Reg and Feed Reg Byp valves 3 will-be closed at 70% Wide Range level. If closing the Feed Reg and Feed Reg Byp. valves does not stop water from go ng into the S/G, P-14 (High High S/G Level) will give a feedwater isolation signal l tripping the operating CF pump. S/G chemistry will be maintained by being lined up to vacuum. Feed line water hammers are being minimized by opening ,
the Feed Reg Byp valve prior'to opening the Feed Ret valve. -
The ND System will be in operation providing core ,
cooling. The S/G will not be used as a heat sink for the NC System during this verification.
PT/2/A/4400/06A NS Heat Exchanger 2A Heat Capacity Test: This test -
determines the shell side fouling factor of the NS l l
. ._.1.b_
e
t 4 ,
Heat Exchanger. _The NS Pump is run in recirculation -i
~
to the FWST, and RN flow is set up"to the maximum a o allowable' amount through'the shell side of the NS' Heat Exchanger. Inlet and outlet temperatures on -
-both the inlet and outlet of the heat. exchanger are measured, along with both the RN and NS flow. Data- :
is-taken using a dLOG data acquisition computer system. All. programs and' calculation used within -t the dLOG computer system are benchmarked and con-g trolled.: Data is analyzed using a Design Engineer-ing Program which has been verified and approved by 3
Design Engineering. 4 There will be no significant effect on the plant .
because_of this procedure. The train of NS under- t test.is entered into the Tech. Spec. Logbook as i inoperable due to the NS Spray Header valves being tagged closed. The other train of NS will be operable at this time.- The procedure cautions the >;
test coordinator'that FWST temperature must be between 70 and 100'F at all times, and the procedure a has the flexibility ta isolate RN to_the NS Heat i Exchanger.immediately if problems are encountered which could lead to an overcooling of the FWST.
i PT/2/A/4400/06B NS Heat Exchanger 2B Heat Capacity Test: This test determines the shell side fouling factor of the. NS -1 Heat Exchanger. The NS Pump is run in recirculation ,
to the FWST, and RN flow is set up to the' maximum allowable amount through the shell side of the NS Heat Exchanger. Inlet and outlet temperatures on both the inlet and outlet of the heat exchanger are measured, along with both the RN and NS flow. Data ,
is taken using a dLOG data acquisition computer system. All programs and calculation used within the dLOG computer system are benchmarked and con- ,
trolled. Data is, analyzed using a Design Engineer-ing Program which has been verified and approved by Design Engineering-There will be no significant effect on the plant because of this procedure. The procedure-cautions the test coordinator that FWST temperature must be between 70 and 100 F at all times, and the procedure j has the flexibility to isolate RN to the NS Heat Exchanger immediately if problems are encountered i which could lead'to an overcooling of the FWST.
PT/1/A/4200/31A S/G PORV Stroke Test: This 10CFR50.59 evaluation applies to PT/1/A/4200/31A, S/G PORV Stroke Test, which was written to demonstrate the ability of the S/G PORV's to ope, against maximum expected main steam pressure (1175 psig). Either actuator pres-sure will be adjusted to account for the difference 119 1
f
+
4~ . l t
i between available steam pressure and maximum ex-pected steam pressure, or' actuator differential q pressure will be verified to.be less than a value ;
which is calculated in the procedure. The calcu- ]
, lated value assures that the valve would open against maximum expected steam pressure.
The only change made to the' procedure since the last 1 issuance of the procedure is in the method by which' .
the valve is stroked to the 30% open position. The procedure now requires that the potentiometer be set at 3 prior to taking the valve to manual control.
The valve will now start travelling to the 30% open position as soon as the control switch is taken to MANUAL. The test results in the past could be influenced by the speed at which the control room -
operator turned the potentiometer to 3 after taking .l the valve to manual control. More consistent ,f results will be obtained with the new method.. -)
i The probability of an accident previously evaluated j
.in.the FSAR will not be increased by this test. The !
probability of the accident in Section 15.1.4,. '!
Inadvertent Opening of a Steam Generator Relief or j Safety Valve, will not be increased since the test 1 involves a " controlled" opening of the valve, not an 1
" inadvertent" opening of the PORV. The control room ;
will have control of the valve at all times, and the ;j valve would be able to close on a-Main Steam Isola- i tion signal. The procedure contains a precaution 'j instruction the operator to close the PORV and/or
~
block valve if steam' pressure decreases to 900 psi or if there .is a dramatic decrease in steam pressure ;
(> 5 psi per second for 5 set.onds). The conse- 1 quences of any accident-previously-evaluated in the FSAR will not be increased by the performance of the test since a stuck open PORV is analyzed in.Section 15.1.4. The possibility of an accident different '.
than any already evaluated in the FSAR is not judged l to be created by this test. A stuck open PORV is i analyzed in Section 15.1.4 of the FSAR, and the purpose of this test is to demonstrate the ability !
of the PORV to open. The probability .of a malfunc-tion of equipment important to safety will not be created by this test. This' test is being performed .
to demonstrate the ability of the PORV.to open. The consequences of a malfunction of equipment important to safety will not be increased by this test. If the PORV were to malfunction, the control room operators can close the block valve. In addition, an open PORV or safety valve is analyzed in Section 15.1.4. The margin of safety as defined in the bases to Tech. Specs. will not be reduced by this test. Prior to opening the PORV, secondary thermal 120 o
p
~
R
,:4 power will be verified by procedure to be less than or equal to 97% due to slight increase (<3%)'that-will occur when the PORV is opened. 'A temporary
' test transmitter is installed on the positioner--
output for testing. This transmitter measures the differential pressure across the valve actuator.
This transmitter will not-adversely affect the ability of the valve to operate. The transmitter is independently verified to be removed and the positioner returned to normal prior to returning the valve.to operable status.
1 PT/1/A/4150/13B Calorimetric Reactor Coolant Flow Measurement: The purpose'of,PT/1/A/4150/13B is to determine the total flow of all Reactor Coolant Loops and to normalize the elbow tap instrumentation. Neither-the.proba-- 1 1'.
bility nor the consequences of an accident will be !
increased by this test. The unit will be in a 1 normal operating alignment with the exception of the l following:
]
- 1) CA tempering will be isolated i
- 2) S/G blowdown will be isolated, and
- 3) NM~ sampling of the NC system will be isolated ;
Tempering flow is provided to the CA nozzles:to cool 'l the inner surfaces of the nozzles and adjacent connected piping and to maintain the water tempera-ture in the piping connecting to the nozzles at ;
approximately feedwater temperature, which should 1 cause the thermal stresses induced in the nozzles and connecting pipe to be reduced when main -j feedwater temperature flow is transferred to the d '
auxiliary nozzles. Each test run should.take 20 minutes with the three runs required by the test 1
-taking between one and two hours. The CA nozzle .j temperatures should not approach the alarm limit during the test; however, action can be taken to cool the nozzles if required. S/G chemistry will be affected by the isolation of blowdown, but it will '
not present any problems due to the short duration of the test. Isolation of NM sampling from NC also .:
will not present problems due to short duration of- 1 test. ;
The performance of this test will not increase the probability or consequences of a malfunction of .j equipaent important to safety. The unit will not be placed in any unusual operating alignments other than those explained above. In addition, the installed test equipment will not increase the ;
probability of a malfunction of equipment. The NC :
hot and cold leg temperatures will be recorded from the spare RTD's. CF temperatures will be recorded 1 21 1
m 3 from test RTD's installed in test thermowells. CF flow and pressure will be recorded.from test in- i strumentation installed on the spare tap sets on the. '
feedwater venturis. The SM pressures will be
- recorded from test. instrumentation installed on the instrument loops which control the S/G PORV's, but actuation of the PORV's will not occur since the transmitter would have to sense a high SM pressure. ,
The transmitter may sense a low pressure while the instrument is being installed. . In addition, Control' ,
Room is notified prior to installation of equipment.
The margin of safety as defined.in the bases to Technical Specificatiens will not be reduced by.the performance of this test. ,
PT/1/A/4250/03E, CA System Discharge Control Valve Throttling Change #8 Procedure: The purpose of this change is to replace' -
the enclosure for CA Pump #1 flow balancing. The maximum pump speed has been increased from 3620 rpm to 3650 rpm, resulting. in higher performance. The new enclosure reflects valu s for flow balancing provided by Design Engine ning per CNSD-1223.42-00-11, Rev. 3.
The procedure change will not. increase the proba-bility of or consequences of any' accident previously analyzed in the FSAR. The new valves have been calculated by and verified by Design Engineering for acceptable CA flows for postulated accidents. No new accident scenarios are created by this change '
since the test method is not affected.
The possibility of and consequences of equipment malfunctions are not increased. .The flowrates and speed are both all within design. limitations for the ,.
pump. This change does not affect the manner in which the pump is operated.
This change does not make any of the CA System inoperable, nor does it lesson the CA System's ability to mitigate the consequences of an accident.
Therefore, the margin of safety as defined-in the Tech Specs is not reduced.
PT/1/A/4250/130 CA Pump 1B Head Verification: The purpose of this procedure is to perform a precision head-capacity I test on CA Pump 1B. These results will be used for
two purposes. The first is to detect possible pump . I degradation over time for various points on the head I curve. The second is to verify pump performance for l balancing CA flows to the steam generators per PT/1/A/4250/03E, CA System Discharge Control Valve .
122
+ _ __
t r
i Throttling Procedure. The balance is different for various pump strengths.
In order to perform this test, CA Pump.IB will be aligned in recirculation to the Upper Surge Tank-
. (UST) and data will be taken at various flowrates. ,
This test may be performed in any mode.- If.in modes l 4, 5, or 6, the CA System is not required to be >
operable so no impact on safety is created.- If in modes 1,.2, or.3, one CA Pump is allowed ~to be out .,
of service for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per Tech Spec Action- '
3.7.1.2a. One Motor-Driven CA Pump and the Turbine
- Driven Pump will be available during this time ~.
Therefore, neither the probability of nor conse-. ,
quences of an accident pre','iously evaluated in the FSAR will be increased. No new accident scenarios will be created by the performance of this test.
The pump will be run in. a normal test lineup with
- sufficient suction head verified prior to pump start. The pump will also be run at flowrates well within design values. Therefore, neither the possibility of nor consequences of an equipment ,t malfunction are increased. No new equipment m '
fuictio" possibilities are created.
Since the CA System will still have two available pumps and will be in a condition allowed by Techni-cal Specifications, the margin of' safety will not be reduced.
PT/2/A/4350/02E, CA, CF, and Turbine Interlocks Periodic Test: This Change #25 procedure change is being written to allow ESF'
- stroke time testing of the following valves: ,
2CF46, S/G 2C CF- Control 2CF37, S/G 2B CF Control '
2CF48, S/G 2C CF. Bypass Control 2CF39, S/G 2B CF Bypass Control 2CF28, S/G 2A CF Control 2CF55, S/G 20 CF Control 2CF30, S/G 2A CF Bypass Control 2CF57, S/G 20 CF Bypass Control 1
i These valves are required to be stroke time tested t using relay K601 in 2SSPSA and 2SSPSB. K601 relay l plunger will be depressed, which will initiate an !
ESF signal to the above eight valves, both CF pump turbines, and the following CF and CA valves:
2CF51, S/G 2C CF Containment Isolation 1 2CF42, S/G 2B CF Containment Isolation 2CF88, S/G 2C CF-Containment Isolation Bypass l 2CF89, S/G 2B CF Containment Isolation Bypass 123 k,.
. m g
i I'
2CA151, S/G 2C CF Byp to.CA Nozzle 2CA150, S/G_28:CF Byp to CA Nozzle 2CA187, S/G 2C CF Temp to:CA Nozzle 2CA186, S/G 28 CF Temp to CA Nozzle !
2CF33, S/G 2A CF Containment Isolation
~
2CF60, S/G 20 CF Containment Isolation 2CF90, S/G 2A CF Containment Isolation Bypass y 2CF87, S/G 20 CF Containment Isolation. Bypass- ,
hC 2CA149, S/G 2A CF Byp to CA Nozzle .
2CA152, S/G 2D CF Byp to CA Nozzle 2CA185, S/G 2A CF Temp to CA Nozzle 2CA188, S/G 20 CF Temp to CA Nozzle These valves will have a " Required Test Position" of- :
CLOSED and_both CF pumps will be in tripped status.
Thus, the only components to stroke during'the test ,
~
are 2CF46, 2CF37, 2CF48, 2CF39, 2CF28, 2CF55,~2CF30, and 2CF57.
Response time testing will be accomplished by j connecting a set of leads from across terminals 3 !
and 4 of TB649 in-2SSPSA and 2SSPSB.to' digital point 03596, " Manual ESF Initiate". When the plunger of i relay K601 is depressed, digital point 03596 will i start the OAC Response Time Testing program, which '
will stroke timo-the eight CF valves under test.
Auxiliary Safeguards Cabinet relay (2A and 2B) S821 will be used to reset K601 relay af ter stroke time testing-has completed.
The K601 relay is located downstream of any other ESF equipment and therefore manipulation of this relay will not affect other ESF components, except those previously described in this analysis.
For this reason, the probability / consequences of any-new or previously-analyzed accident or equipment-malfunction is not increased, and the margin of- i safety is not reduced.
- 1 PT/1/A/4450/05A, Containment Air Return Fan and Hydrogen Skimmer Fan Change #20 1A Performance Test: Step 12.'1.12 is deleted because damper ARF-D-2 is verified closed and then-the breaker is opened in steps 8.3 and 8.3.1. With the breaker opened, the damper can not move so it is assured to be in the closed position. Step 12.3.12.1 is changed to shut down HSF-1A and Step-12.3.12.2 is added to close valve IVX-1A. The note before Step 12.3.11 was added to ensure that the HSF-1A is shut down quickly, after IVX-1A has opened, and the suction valve (IVX-1A) is closed to reduce the possibility of steam bypassing the ice condenser in the event of an accident (LOCA) during ;
124 e
- w. ,
the test. Deletions from steps 12.3.11 and 12.3.12 -
are covered under the added note, Step 12.3.12.1 and-Step-12.3.12.2.- These changes do not reduce _the ,
margin of safety as defined in Tech Specs, create.or, l increase the probability of an accident or increase the consequences of an accident. Malfunctions of equipment important to safety will not be increased' or created and the consequences of a malfunction will not_ increased. Therefore, an unreviewed safety question does not exist.
OP/1/A/6700/01 Unit One Data Booki The Target AFD in Data Book Figure 1.1 is updated by performance of PT/1/A/4150/08, Target. Flux Difference Calculation.
The target AFD is changed to keep Control Bank D at-approximately 215 steps withdrawn (at 100% Full Power) while allowing-for changes in the natural !
axial power shape that occur with burnup. _ The -
targets are set by procedure te be within the f operating bounds of T.S. 3/4.2.1. The targets-are an operating guideline only to aid the control room.- ;
operators in maintaining AFD within the; limits of T.S. 3/4.2.1. The targets serve no other purpose. -
They do not feed any trip function or serve any safety related functions. The limits'that must be observed in T.S. 3/4.2.1 are set by cycle specific '
analysis.
7 AFD is an input to the OTDT Trip Setpoint. Targets may be set so that the AFD input-function to the-OTDT will impose a penalty on OTDT. However, this penalty will automatically be imposed by the 7300- a system per the formulation of T.S. 2.2.1. ;
AFD anC AFD targets are monitored by the OAC NUCLEAR ,
06 (AFD ALARM MONITOR) program. AFD Targets on the OAC are changed by the above mentioned procedure,.
also.
All accidents analyzed in FSAR Chapter 15 have as
- one of the initial conditions that AFD is within the limits of T.S. 3/4.2.1. As such the targets do not !
have any effect on the accident analysis.
l MP/0/A/7150/39 Reactor Coolant Pump Seal Removal and Replacement:
This safety evaluation is for the reissuing of MP/0/A/7150/39 changes 0 to 12 incorporated, pre-4 pared on 01-31-1989. This procedure has been completely rewritten. Since this procedure has been completely rewritten, this safety evaluation will address the procedure as a whola.
A full USQ evaluation is required because this procedure was changed significantly and it is 125
'I
m y P
described in the FSARJ FSAR Section' 13.5.2.2.1 (Maintenance Procedures) states that maintenance i procedures are required for safety related equip-F, ment.- Since this rewrite, as described above, changes this procedure in a significant manner, a:
full USQ evaluation must be performed.
Tech Specs 3.2.3, 3.4.1, and 5.4 are affected by
~
this' procedure. Operations has the responsibility and'the procedures-for compliance with these Tech Specs. Maintenance will be performed on this pump when Tech-Specs allow, per Operation's procedures.
This rewrite will clarify and assure that mainte-nance activities will return the pumps to as-designed conditions.
The changes made by this rewrite have been reviewed against approved vendor manuals, design documents, j and station procedures to ensure that the corrective maintenance controlled by this procedure will return the pump to as-built /as-designed condition. These- -
actions will ensure the pumps' compliance with FSAR accident analysis. Since the pump will be returned }
to as-designed conditions, the possibility, conse-quences, or probability of a malfunction will be reduced. Therefore, no USQ exists.
- PT/2/A/4200/01T Containment Penetration Valve Injection Water System "
Performance Test: The Containment Penetration Valve Injection Water System (NW) is tested by venting the Containment Isolation Valve (s) under test on both sides of the valve-seat (s) and then measuring the NW flow to that (those). Containnient Isolation Valve (s). .
Containment Isolation Valves (CIVs) may be NW. System-leak tested in groups or individually.
This test is designed primarily to be conducted when containment integrity is not required (Mode 5 or-when fuel is totally removed from the core - No-Mode). The test may, however, be performed on an.
as-need basis with the prior approval of Operations, j When testing in Modes 5, 6, or No Mode, the initial alignment of hand-operated NW isolation valves-from-Operational Readiness condition to Test Condition ,!
will be performed by Operations per OP/2/A/6200/19.
Testing will normally proceed for an extended period of time, involving many other hand-operated NW L isolation valves' alignments. These alignments will be performed by Performance Technicians. Once all
-desired testing is complete, the test procedure ,
requires Operations to restore the NW System to full
Operational Readiness per OP/2/A/6200/19.
126
- g. .
a; i !
The NW System Leak-Test may also be performed in Modes 1 thru.4, provided the test.is coordinated with Operations regarding Train operability of NW AND the Containment Isolation Valves which must have. ;
s their NW (Seal Water) isolated in order to perform -l the test. By isolating NW to any CIV, that CIV must be declared inoperable and logged into the Technical-Specifications Action Item Log (TSAIL) as inopera-ble. As a result _of the necessity for such strin- .
gent precautions and controls, NW System testing in Modes 1-4 will only be attempted if continued Unit i operation is a concern.
This test does affect NW structures, systems, and. [
components addressed in the FSAR in a significant ;
manner. But by testing when NW operability is not :
required (Modes 5, No Mode) and by testing in Modes 1-4,-and 6 with Operations approval (i.e. Operabil-- -1 ity Determination for Modes.1-4, Containment Integ-rity-Determination),_no significant challenges to the Tech Spec required function of the NW System are presented. In addition, Operations' control of the o Operational Readiness of the NW System through use of PT/2/A/6200/19 will ensure that the NW System-is fully available'when required.
MP/0/A/7600/27A, Borg Warner Seal. Welded Handwheel Operated Gate
- Change =#2 Valve Corrective Maintenance: This procedure change '
decreases the packing retainer torque:for the. valves under repair. . Previous values were set higher than ,
required. The valves are handwheel operated and, therefore, the higher stem friction caused by the higher torque will not adversely-affect plant ,
operations. The lower torque will improve packing life.
The Catawba FSAR and Technical Specifications have been reviewed and are not affected by this procedure _'
change. This procedure applies to valves used in various plant applications. The safety standards previously specified within the FSAR and Technical Specifications will not be af fected by this proce-dure change. The probability.of an accident or
, m - malfunction previously addressed in the FSAR will not be increased. No unreviewed safety questions ,
are involved.
OP/1/A/6700/01 Unit One Data Book: The Target AFD in Data Book Figure 1.1 is updated by perfctmance of PT/1/A/4150/08, Target Flux Difference Calculation.
The target AFD is changed to keep Control Bank 0 at approximately 215 steps withdrawn (at 100% Full Power) while allowing for changes in the natural axial power shape that occur with burnup. The 12.2 j
- ' h targets are-set by procedure _to be within the operating bounds of T.S. 3/4.2.1. The targets are- l an operating guideline only to aid the control room 1 operators in maintaining AFD within the limits of T.S. 3/4.2.1. The targets serve no other purpose.
They do not feed any trip function or serve any ,
- =fety related functions. The limits that must be observed in T.S. 3/4.2.1 are set.by cycle specific ;
analysis. .
i AFD is an input to the OTDT Trip Setpoint. Targets may be set so that the AFD input function to the i OTDT will impose a penalty on OTDT. However, this '
penalty will-automatically be imposed by the 7300 "
system per.the formulation of T.S. 2.2.1.
AFD and AFD-targets are monitored by the OAC NUCLEAR 06 (AFD ALARM MONITOR) program. AFD Targets on_the c OAC are changed by the above_ mentioned procedure, i also.
All accidents analyzed in FSAR Chapter 15 have as one of the initial conditions that AFD is within the limits of T.S. 3/4.2.1. As such, the targets do not have any effect on the accident analysis.
OP/1/A/6700/01, Unit One Data Book: This procedure change removes
' Change #150 the- temporary rod withdrawal limits (Data Book Curve 1.2.1) incorporated via pro:edure change No. 143,-
(* which were generated per PT/1/A/4150/20, Temporary Rod. Withdrawal Limits Determination. The purpose of these limits was.the imposition of operational limitationsfon control rod withdrawal to ensure that the reactor coolant was maintained at critical boron- ,
concentrations low enough for Moderator ~ Temperature :
Coefficient to be in Compliance with Tech Spec 3.1.1.3. Based upon evaluation of Core Design Data (Unit.1, Cycle 14 Startup and Operational Report), it has been determined that the All' Rods Out Critical '
Boron concentration has been reduced enough at a core depletion of 34 EFPD to remove restrictions on control rod withdrawal.(i.e. NCS critical boron ,
concentration is low enough with all- control rods withdrawn to be assured of being in compliance with
- the Tech Spec restrictions on MTC).
Neither the probability nor the consequences of any -
accident or safety significant equipment malfunc-tion, either analyzedlor unanalyzed by the FSAR, will be increased by this change. This assures continued compliance with assumed valves of MTC (used in the reviewed FSAR analyses) through com-pliance with Tech Spec 3.1.1.3. The margin of safety as defined in the applicable Tech Spec bases
p- .
- , 1 p -,
t i
is not reduced due to full compliance with Tech g Specs.
?- TT/0/A/9100/45 .RN 2 Out of 3 Pond Swap: This procedure will verify the 2 out of 3 swap over logic-for RN to the SNSWP. =
Each level transmitter will be _ individually checked C to ensure that.-a signal from a single transmitter cannot initiate a swap over to the SNSWP. After this is done, a 2 out of-2 signal will be. simulated, e causing a swap over to the:SNSWP. The signal will
. be cleared, and the other signal combinations will' be tested while in the swapped over alignment. .
s Relays will be verified to verify that all combina-
- tions of the-swap over logic are correct.
When the pond swap is initiated, train separation of !
the RN system will occur, RN will align to the SNSWP and all four RN pumps will start. Since the SNSWP '
1:, the assured source of RN, alignment to the pond-T is conservative. The procedure provides for Opera-
tions to use the NS heat exchangers to provide _
miniflow for the RN pumps during the time when all four RN pumps are operating so there will be no adverse effect on the RN pumps, l
MP/0/A/7200/05 Auxiliary Feedwater Pump Turbine Corrective Mainte-nance: This safety evaluation is for the re-write of MP/0/A/7200/05.- The following is a summary of
- changes made to this procedure. 1 Upgrade of Section 2.0 through 10.0 to include more references, pre-requisites, special tools, and acceptance requirements.
Disassembly of turbine; Sections 11.2 - 11.14 were re-written to provide greater detail in ,
description of parts and inspection' procedure in accordance with the approved manufacturer manual. Inspection now includes taking rotor run-out and thrust "as-found" during disassem-bly. ,
i' A hold point was added prior to rotor disas-sembly because CNS stocks a complete rotor assembly. This must be evaluated by M.E.S.
accountable engineer prior to proceeding further.
A step and hold point was added to perform magnetic particle inspection on turbine rotor
- assembly to check for cracks and defects. This L is in response to PIR 87-0371. [
l 1
Y
h >
F n
0 -
Assembly of rotor and turbine, Sections 11.55-11.23 were also re-written to. include more detailed descriptions and assembly practices to-minimize error and clarify work progression.
Enclosure 13.1 was revised to reflect changes in the procedure body:for added hold points a d inspection results documentation. ] .
Enclosures 13.2 through 13.4 were added to include drawings of the turbine.
Enclosure 13.6 was expanded to include the recommended torque values and procedures.
-A full USQ evaluation is required because this procedure is being changed significantly and is !
addressed in the FSAR. FSAR Section 13.5.2.2.1 a states that maintenance procedures are required and .;
since this re-write, as described above,-changes
- this procedure in a significant manner, a USQ evalu-ation must be performed.
Technical Specification 3/4.7.1 is affected by this procedure, Operations has the responsibility and the procedures for compliance with this Tech Spec.
Maintenance will be performed on this turbine when Tech Specs allow, per Operations procedures. The .
changes made by this re-write will not affect this Tech Spec since-these changes will clarify the 3 l procedure and assure that maintenance activities will return the turbine to as-designed conditions. .
The changes made to this procedure have been re-i.
viewed against approved vendor manuals, design l documents, and station procedures to ensure that the corrective maintenance _ controlled by this procedure will return the turbine to as-built /as-designed <
condition. These actions ensure compliance with the FSAR accident analysis. Since the turbine will be returned to as-designed conditions, the possibility, consequences, or probability of a malfunction will l be reduced.
OP/1/A/6700/01, Unit One ' Data Book: The. Target AFD in Data Book
' Change #147 Figure 1.1 is updated by performance of PT/1/A/4150/08, Target Flux Difference-Calculation.
l The target AFD is changed to keep Control-Bank D at ;
approximately 215 steps withdrawn (at 100% Full Power) while allowing for changes in the natural l- axial power shape that occur with burnup. The l targets are set by procedure to be within the operating bounds of T.S. 3/4.2.1. The targets are an operating guideline only to aid the control rcom 130 l o
P" 4
J L
-operators'in maintaining AFD within the limits _of T.S. 3/4.2.1. The targets serve no other purpose, b, _ They do not feed any trip-function or serve any safety related functions. The limits that must be -
observed in T.S. 3/4.2.1 are set by cycle specific analysis. !
AFD is an input to the OTDT Trip Setpoint. Targets may be set so that the AFD input function to the -
OTDT will impose a penalty on 0 TDT. However, this penalty will automatically be imposed by the 7300 system per the formulation of T.S. 2.2.1.
AFD and AFD targets are monitored by the OAC NV_ CLEAR [
06 (AFD ALARM MONITOR) . program. AFD Targets on the OAC are changed by the abcve mentioned procedure, '
also.
All accidents analyzed in FSAR Chapter 15 have as-one of the initial conditions that AFD is within the limits of T.S. 3/4.2.1. As such, the targets do not have any effect on the accident analysis.
PT/1/A/4250/03E, CA System Discharge Control Valve-Throttling
' Change #9 Procedure: The purpose of this change is to modify:
the step for visual stroke timing of the CA Control Valves if they are throttled during_the test.
Article IWV-3200 requires that a retest be performad if a valve'or its control system is replaced or had maintenance performed that could affect its nerfor-mance prior to returning the valve to service. Past experience has shown (see data below) that stroke times typically vary from 1 0.4 to.1 0.8 seconds between each test. The stroke time variation following stem adjustment of one thread is less than-or equal to-the Lvariation normally seen between tests. Therefore, adjusting the stem travel less-than or equal to 1 thread does not affect the valve performance. In addition, any adjustment to close l the valve any amount would result in a shorter stroke time, which would be in the conservative direction. After the flow balance was completed, an-.
engineering judgement would be made'by either the CA System Expert or the IWV Program Expert to determine if'a visual stroke time test is necessary. A work-request is normally written after any adjustment of these' valves to ensure that the limit switches are set up' correctly. A stroke time test will be performed following this work.
i 131
V b.
7 G
Typical Date. Old Old New Visual.New . Time Valve Time Diff_ Throttled Pos Time Pos Time Time Diff
-ICA56 1 0.8- 8/27/88 11 3.4 12 3.7 3.4 0 $
-i 1CA60 1 0.8 8/27/88 10 3.4 11 3.8- 3.6 +0.2 ,
ICA44 1 0.6 8/27/88 10 4.0 12 4.1 4.2 +0.2 1CA40 1 0.6 8/27/88 12 4.6 13 5.0. 5.2 +0.6 r L ICA52 ;1 0.4 8/27/88 7 2.6 8 3.4 3.4 +0.8 -
Notes: - Position is number of threads on ; tem exposed -'
- Times are in seconds
- Stroke time difference is between vid stroke time and new 1 stroke time with limit switches adjusted ti i
Since this change only affects whether or not a '
visual stroke time has to be performed, the proba-bility of and consequences of L:cidents previously evaluated in the FSAR are not increased.- In addi-tion, no new accident scenarios are creat'ed by this change since there is a negligible ~ stroke time dif ference af ter a one thread adjustment and the stroke times are well below the maximum stroke time of 20 seconds per IWP.
The valves are being throttled-to provide flow rates well within the design limits.of the CA pumps. This change does not' affect the actual throttling of
-these valves. Therefore,-there is no effect on equipment malfunction possibilities,. probabilities, .,
or consequences.
Since this change does not affect the actual test i method, the margin of safety as defined in the bases of the Technical' Specifications is not reduced.
MP/0/A/7400/09 Diesel Engine Cylinder Head Removal.and Replacement:
This is a previously reviewed and fully approved procedure that a procedure step .is being added.
Added Step 11.3.35.2 to provide procedure step to replace or install a new adjusting screw assembly '
when a rocker is replaced or the assembly is worn, thereby restoring the component to original conditions. The addition does not create an unreviewed safety or detract from operability. It does not effect any section referenced and in a significant manner.
132 l
~
p r
..p PT/2/A/4200/01C Containment Isolation Valve Leak Rate Test:
r PT/2/A/4200/01C, Containment Isolation Valve Leak Rate Test is performed to measure the leak rates of
, , all containment isolation valves required by Catawba FSAR Table 6.2.4-1 and to satisfy Technical Speci-fication Surveillance requirements. The test requires draining and/or venting of the penetrations ,
to be tested per OP/2/A/6200/20. A volumetric leak -
rate monitor is then' connected to the penetration,
the penetration is pressurized and the leak rate is measured.
The only changes to this procedure involve altering E the valve lineups to satisfy containment closure requirements. Containment closure is not violated for any penetration during the performance of this procedure. Closure is always assured by either the inboard or outboard containment isolation valve
.(CIV) and applicable vents and drains. When 4 OP/2/A/6200/20 is finished draining / venting the penetration, it has both CIVs and any vents or drains between them closed. The lineup for tnis test is then performed which maintains closure on one side of the penetration. When the' test is finished, the penetration is left with both CIVs and any vents and drains between them closed.
The above mentioned changes altered the valve 'I lineups where needed to ensure containment closure.
The validity of the type C test is not compromised.
There were no changes that would alter the leak rates measured by this procedure.
Therefore, the probability and consequences.of an
' accident or malfunction of equipment important to l safety previously evaluated in the FSAR will not'be ,
increased. The possibility of an accident or malfunction of equipment important to safety not previously evaluated in the FSAR will not be creat-k ed. The margin of safety in the basis to' Tech Specs 1 will not be reduced. ,
o EP/1/A/5000/10 Steam Line Break Outside Containment; The signifi- !
car,t change to this procedure is the deletion of Step 2.6 which was a redundant check for a faulted S/G. The RNO of this step was used to check for an isolated steam line break downstream of the MSIVS. '
EP/01, reactor trip or safety injection, has been changed under retype #8 so that the flowpath for terminating S/I after an isolated steam line break is no longer through this procedure, but through
, EP/18, S/I termination following spurious S/I. It is felt that an isolated steam line break can be treated like a spurious S/I since the cooldown is quickly terminated due to isolation of the break by 133
t faulted S/G. See' EP/01 retype #8 10CFR50.59 evalu-ation for justification of this change.
Since the flowpath for responding to an isolated steam line break has changed, this deleted step .is no longer needed in this procedure since it only
~
applies for the. isolated steam line break scenario, s This change will require a change to the' Catawba-EPGS. Design Engineering-(Safety Analysis Group) has been notified and. agrees with this change and is
~
pursuing the change to the Catawba EPGS. Since this change does not affect the evaluation of steam line-
~
breaks as given in the FSAR.nor does it create the F ' possibility of an-accident different from that evaluated in the FSAR, it is felt that this change ;
does not represent an unreviewed safety question. '
MP/0/A/7600/24 Pacific Pressure Seal Tilting Disc Check Valve Corrective Maintenance: This procedure performs main steam isolation. This is different from an unisolated steam line break where and extended ;
cooldown exists due to the complete blowdown of the-corrective maintenance on Pacific pressure seal tilting disc check valves. Instruction manual CNM- 1 1205.00-0318 provided technical information in the {
development of the procedure. Plant specific :
information that applies to Catawba Nuclear Station was also included in its development.
This evaluation is for changes made during the _
procedure upgrade process. These changes are in the !
procedure format and are not significant in content.
The_ Catawba FSAR and Technical Specifications have-been reviewed and are not affected by this proce- ,
dure. This procedure applies to valves used in various plant applications. This procedure will be ,
used to correct and improve the performance of the. -
valve and will maintain the valve within its origi-nal design requirements and specifications. The safety standards previously specified within the 1 FSAR and Technical Specifications will be main-tained. The probability of an accident or malfunc-tion previously addressed in the FSAR will not be-increased. No unreviewed safety questions are involved. ,
y MP/0/A/7600/80 Kahn Process Check Valve Corrective Mair.tenance:
This procedure performs corrective maintenance on Kahn process check valves.
This evaluation is for changes made during the procedure upgrade process. These changes are in the procedure format and are not significant in content.
This procedure and the change will be used to correct and improve the performance of the valve and 134_
b b
y wi11' maintain the valve within its-original design-requirements and specifications. %e Catawba FSAR and Technical Specifications have been reviewed and j are not affected by this procedure.. This procedure- i applies to valves used in various plant applica- i tions. This procedure will maintain the valves in all applications to the safety standards specified )
within the Catawba FSAR Md Technical.Specifica- '
tions. The probability of an. accident or a mal-function previously addressed in the FSAR will not <
be increased. No new unreviewed safety questions q are involved, i PT/2/A/4550/09 Shuffle Verification: This procedure is designed to.
verify that all fuel assemblies that are going back into the core for Cycle 3 are in the proper SFP locations and contain the proper Cycle 3 inserts.
To do this, each fuel assembly is inspected and its f location, RRN and insert ID recorded. Then this record is independently compared to the Af ter ;
Shuffle map found in PT/2/A/4150/18, Fuel Assem- "
bly/ Insert Shuffle Procedure. If any discrepancies are found, they are resolved before completion of li the procedure. .;
This procedure is designed to prevent the accident i analyzed in FSAR Chapter 15.4.7, Inadvertent Loading and Operation:of a Fuel Assembly in an Improper Position. Further prevention of this is provided by PT/2/A/4550/03C, Post Reloading Core Verification, l:
after Core Loading is complete. ]
While loads are transported across the spent fuel in' ,
the SFP, the-loads consist of only an underwater ~j camera, support poles and cable. This is much less 1 than the weight of a fuel assembly and is.bnunded by l the accident analysis in FSAR' Chapter 15.7.4, Fuel q Handling Accidents. 1 u ;
This procedure does not require off-normal operation !
of safety ' equipment so there will be no increase in ,
the probability or consequences of a malfunction of j equipment important to safety .as previously evalu- i ated in the FSAR nor create a possibility of equip- ;
ment malfunction important to safety different than 1 any evaluated in the FSAR. l J
No safety features are jeopardized and no Tech Spec .
requirements are violated by this procedure, there- 3 fore, the margin of safety is not reduced.
MP/0/A/7600/80 Kahn Process Check Valve Corrective Maintenance: <
This procedure performs corrective maintenance on Kahn process check valves.
JJ35 l
a.
]
q This evaluation is for changes made during the procedure upgrade process. . These changes are in the.
procedure format and are-not significant in content. -
This procedure-and the change wil_1 be used to correct and improve the performance of the valve and will maintain the valve within,its original design requirements and specifications ~The Catawba FSAR and Technical Specifications have been reviewed and-are not affected by this procedure. This procedure applies to valves used in various plant applica-tions. This procedure will maintain the valves in.
all applications to the safety standards specified '
within the Catawba FSAR and Technical Specifica-tions. The probability of an accident or a mal-function previously addressed in the FSAR will not be increased. No new unreviewed safety questions i are involved. ,
PT/1/A/4600/11, Neutron Noise Data Acquisition: This change' adds r Change'#6 provisions for the monitoring-of three additional Incore Thermocouples (subject procedure presently allows for the monitoring of seven of the Incore-T/C's) for the purpose of assessing-the behavior of observed reactor coolant flow anomaly. None of these T/C's are Safety Grade, so. Tech Spec 3.3.3.6 is not affected by-this change. Indication of core exit temperatures by these thermocouples will not be affected by visicorder trending via test leads installed in the ENA Cabinet in the Control Room.
0AC Program Nuclear 11, Thermocouple Map, will still o receive valid process indications for the purpose of calculating core quadrant power tilt and FAH as ,
inferred by T/C indications. The basis for the calculations of the Thermocouple Map-program is stated in FSAR Section 4~4.6.1 which refers to the Incore T/C's as a backup for the Incore Flux Mapping .
System.
Neither the probability nor the consequences of any accident or safety significant equipment malfunc-tion, either analyzed or unanalyzed by the FSAR, will be created by this change. The Incore T/C's themselves (described in FSAR Section 7.7.1.9.1) are-not.affected in any way by this change. The margin of safety of Tech Spec will not be reduced since this change has no impact on any Tech Specs.
The installation of another Scaler-Timer does not present an unreviewed Safety question or require a change to the Technical Specifications. This is true since the Scaler-Timer will only be used for data gathering.
136
y ir y
This change is intended for lower the probability that obtaining Increase Count rate Ratios (ICRR) data will be a critical path item during refueling.
The temporarily installed scaler-timer will be identical to the permanently installed one. .It will be installed in the same manner as the permanent scaler-timer is installed. Signals to the Audio Count rate circuit will not be affected.
Instrumentation requirements of Tech Specs will not be affected by change.
PT/2/A/4200/01C, Containment-Isolation Valve Leak Rate Test:
Change #48 Restricted Change #48 of PT/2/A/4200/010 is being written to allow testing 2NM25A and 2NM425 as a test separate from other valves of Enclosure 13.14 (Penetration M310), i.e. 2NM22A, 2NM26B. Enclosure 13.14 is currently arranged so that all NM valves are tested in consecutive order--2NM22A, 2NM25A and 2NM425, 2NM268. .l 1
Enclosure 13.14 has already been performed once during U2-E002. During that test, high leakage was identified attributable to 2NM25A and 2NM425. This change is therefore written to allow a retest of 2NM25A and 2NM425~ separate from 2NM22A and 2NM260.
Since this procedure change does not change the test I1 alignment or procedure for testing 2NM25A and j' 2NM425, and only deletes other sections which have already been successfully tested, neither the probability nor the consequences of an accident ;
previously evaluated in the FSAR will be increased. 1 No new-accident scenarios will be created by the i
! performance of this test. In addition, neither the !
L possibility of nor consequences-of an equipment.
I' malfunction are increased. No new equipment mal- ;
L function possibilities are created. lg PT/1/A/4550/01E Spent Fuel Building Manipulator Crane: ;
PT/1/A/4550/01E was originally the load test portion j of PT/1/A/4550/01A (Preparation for Refueling). It j was separated from PT/1/A/4550/01A (Preparation for i Refueling) to improve useability of the procedure- .
and to allow simplified retesting of the mast j following maintenance to it. Additional changes to 'l the content have been made from the original proce-dure to ensure the operator understands the event to :
take place following an action prior to the operator performing the action. PT/1/A/4550/01A gave fixed '
numbers for the Overload, Underload and Lowload setpoints of the Crane. These settings are no l 1
longer fixed, but are calculated in OP/1/A/6550/06 I
137 w
m ,
q (Transferring Fuel with the Spent Fuel Manipulator Crane) based on actual weights of Mast and if needed, an irradiated fuel assembly. The above calculations are based on Westinghouse F5.0 and F5.1 specifications to ensure the safe movement of fuel.
The mechanical overload setpoint has been reduced l from 3250 lbs i 20 lbs to 2540 2 160 lbs to allow a more conservative setting cf this back up device's
, setting. The allowable span was increased to allow a more realistic operating band for this mechanical.
See attached letter on' requested FSAR update to FSAR Section 9.1.4.
The development and use of this procedure does not change probability or consequences of any accident or equipment malfunction discussed in the FSAR. Nor i
does it create the possibility of a different accident or equipment failure scenarios. Load testing of the Spent Fuel Building Manipulator Crane-is not required by Tech Spec and, therefore, the Tech Spec Bases are not affected by this procedure.
PT/1/A/4550/01A, Preparation for Refueling: PT/1/A/4550/01A retype Retype #3 #3 incorporates changes thru along with additional changes. This retype removes load testing of the Fuel Hoist portion of this PT to PT/1/A/4;50/01E (Spent Fuel Building Manipulator- ;
Crane Load Test) and the New Fuel and New RCCA Handling Tools to PT/1/A/45SO/01F (Preparation for New Fuel b eeipt). Numerous editorial changes have been made throughout the procedure. The major change now directs the operator to the applicable equipment opd ating procedures to perform the various verifications made. This ensures that as i the equipment procedures are updated, the need to update this PT is reduced. It also ensures consistency in the operation et the equipment-and ensure the Limits and Precautions for the applicable equipment are reviewed. Other editorial revisions ensure the operator understands the event to take place following an action prior to the_ operator performing the action.
The performance of this procedure ensures the intent of paragraph 9.1.4.4 of the FSAR is met for the-equipment in the Spent Fuel Pool. The use of this procedure does not change the probability or conse-quences of any accident or equipment malfunction discussed in the FSAR. Nor does it create the possibility of a different accident or equipment malfunction scenario. There are no Technical Specification Requirements for this equipment; therefore, the margin of Safety for the Tech Spec as defined by the Bases is unaffected.
1 33 4 i
. . . . . - . .. . . . . . . . . . . . i
___.___m . - . .. .
PT/1/A/4200/31A S/G PORV Stroke Test: This 10CFR50.59 evaluation applies to PT/1/A/4200/31A, S/G PORV Stroke Test, which was written to demonstrate the ability of the S/G PORV's te open against maximum expected main steam pressure (1175 psig). Either actuator pres-sure will be adjusted to account for the difference between available steam pressure an maximum expected steam pressure, or actuator differential pressure will be verified to be less than a value which is calculated in the procedure. The calculated value assures that the valve would open against maximum expected steam pressure.
The probability of an accident previously evaluated in the FSAR will not be increased by this test. The probability of the accident in Section 15.1.4, Inadvert m Opening of a Steam Generator Relief or Safety Valve, will not be increased since the test involves a " controlled" opening of the valve, not an
" inadvertent" opening of the PORV. The control room will have control of the valve at all times, and the valve would be able to close on a Main Steam Isola-tion signal. The procedure contains a precaution instructing the operator to close the PORV and/or block valve if steam pressure decreases to 900 psi or if there is a dramatic decrease in steam pressure
(> 5 psi per second for 5 seconds). The conse-quences of any accident previously evaluated in the FSAR will not be increased by the performance of the test since a stuck open PORV is analyzed in Section 15.1.4. The possibility of an accident different than any already evaluated in the FSAR is not judged to be created by this test. A stuck open PORV is analyzed in Section 15.1,4 of the FSAR, and the purpose of this test is to demonstrate the ability of the PORV to open. The probability of a malfunc-tion of equipment important to safety will not be created by this test. This test is being performed to demonstrate the ability of the PORV to open. The consequences of a malfunction of equipment important to safety will not be increased by this test. If the PORV were to malfunction, the control room operators can close the block valve. In addition, an open PORV or safety valve is analyzed in Section 15.1.4. The margin of safety as defined in the bases to Tech. Specs. will not be reduced by this test. Prior to opening the PORV, secondary thermal power will be verified by procedure to be less than or equal to 97% due to slight increase (< 3%) that will occur when the PORV is opened. A temporary test transmitter is installed on the positioner output for testing. This transmitter measures the differential pressure across the valve actuator.
This transmitter will not adversely affect the 1M
I p <
. C' '
N r.
ability of the valve to operate. The transmitter is j independently verified to be removed and the positioner returned to normal prior to returning the
. valve to operable status. .
PT/1/A/4250/03C, fuibi.se Driven Auxiliary Feedwater Pump #1' Perform-Change #39 ance Test: The purpose of this change is to e ure that CA Pump #1 is sufficiently challen fold, ouick start on a quarterly basis.'ged byc,.ange This .
is in response to recommendations made in the E Operational Readiness Review conducted by INPO in y January of 1989. This change will result in the pump not being manually started up in minimum speed l with the excess moisture drains open. Instead, the
,. pump will start up at full speed just as it would on y an emergency start signal.
i
- The basic test method and expected results are not being affected as a result of this change. The rest of the system is not affected since the pump is aligned in recirculation and is isolated from the .i' e safety related portions, The step to ensure the alignment is per the Operations procedure is not l
, necessary since a correct valve lineup is provided on Enclosure 13.5, Therefore, the probability of l and consequences of an accident previously evaluated in the FSAR are not increased. In addition, no new accident scenarios are created since the valve alignment and test method are not revised as a i result of this change. l The pump will be started as it is designed to be on an emergency signal. It will also be run well within the design flowrates and~ speeds. Therefore, the probability of and consequences of a malfunction of equipment important to safety are not increased.
Since no change to pump operating parameters is ];
made, no new possibilities of equipment malfunctions are expected to occur. "
Since no change is made to the procedure as to how many operable CA pumps are in service, the margin of safety as defined in the bases of the Technical Specifications is not reduced.
P1/2/A/4200/13E, CA Valves Inservice Test (QU): This change Change #17 increases the percent increase in valve stroke time where an investigation is initiated and the valve is placed on a monthly testing freqbency, from 25% to 50% for 2CA-56. This valve always strokes in less than 5 seconds and IW requirements calls for valves ,
that stroke in less than 10 seconds to use 50%
increase in stroke time as the threshold. The Tech Spec required stroke time for this valve is not 4
140 j
~-
o n _
, a
,1 : t; 'i [
[:[h[
y1 %
i j
f" 1 e s being changed se no unreviewed safety question exists due to this change.
TT/1/A/9200/56 Retest of 1SV1 -ISV7, ISV13, and ISV19 for NSM 'l
,' CN-50395: This 10CFR50.59 evaluation applies to x TT/1/A/9200/56, Retest of ISV1, ISV7, ISV13, and ,
ISV19, which was written to retest the S/G p0RV's J after the completion of NSM CN-50395. ;
W The probability of an accident previously evaluated in the FSAR will not be increased by this test. . The ;
probability of the accident in Section 15.1.4,. ;
Inadvertent Opening of a Steam Generator Relief or Safety Valve, will riot be increased since the test
, . involves a " controlled" opening of the valve,:not an" Q " inadvertent" opening of the PORV. The control room km will have control of.the valve at all times, and the P valve would be able to close on a Main Steam Isola-tion signale The procedure contains:a precaution instructing the operator to close the PORV and/or block valve if steam pressure decreases to 900 psi or if there is a dramatic decrease in steam pressure
(> 5 psi per second for 5 seconds). The conse-quences of any accident previously evaluated in'the FS.AR will not be increased by the performance of the test since a stuck open PORV is analyzed in Section
? 15.1.4. The possibility of an accident different than any already evaluated in the FSAR is not judged to be created by this test. A stuck open PORV is analyzed in Section 15.1.4 of the FSAR and the purpose of this test is to demonstrate the ability L of the PORV to open after the completion of NSM CN-50395. The probability of a malfunction of-equipment important to safety will not be created by N
this test. This test is being performed to demon-strate the ab1Hty of the PORV to open af ter NSM CN-50395 is' implemented. -The consequences of a malfunction of equipment important to safety will not.be increased by this test. If the p0RV were to.
malfunction, the control room operators can close the block valve. in addition, an open PORV or ,
safety valve _is analyzed in Section 15.1.4. The margin of safety as defined 1n the bases to Tech.
Specs will not be reduced by this test. Prior to opening the p0RV, secondary thermal power will be O. verified by-procedure to be less than or equal to 97% due to slight increase (< 3%) that will occur when the PORV is opened. A temporary test trans-mitter is installed on the positioner output for p testing of ISV13. This transmitter measures the differential pressure across the valve actuator.
This transmitter will not adversely affect the ability of the valve to operate. The transmitter is independently verified to be removed and the Mi T
p V
! pos,tioner returned to normai prior to returning the valve to operable status.
PT/1/A/4200/31A S/G PORV Stroke Test: This 10CFR50.59 evaluation r applies to PT/1/A/4200/31A, S/G PORV Stroke Test, F which was written to demonstrate the ability of the S/G PORV's to open against maximum expected main steam pressure (1175 psig). Actuator pressure will
_ be adjusted to account for the difference between L available steam pressure and maximum expected steam pressure.
The probability of an accident previously evaluated in the FSAR will not be increased by this test. The
- probability of the accident in Section 15.1.4, Inadvertent Opening of a Steam Generator Relief or Safety Valve, will not be increased since the test involves a " controlled" opening of the valve, not an
" inadvertent" opening of the PORV. The control room will have control of the valve at all times, and the valve would be able to close on a Main Steam Isola-tion signal. The procedure contains a precaution instructing the operator to close the PORV and/or block valve if steam pressure decreases to 900 psi or if there is a dramatic decrease in steam pressure
(> 5 psi per second for 5 seconds). The conse-quences of any accident previously ,/aluated in the FSAR will not be increased by the performance of the test since a stuck open PORV is analyzed in Section 15.1.4. The possibility of an accident different than any already evaluated in the FSAR is not judged to be created by this test. A stuck open PORV is analyzed in Section 15.1.4 of the FSAR, and the purpose of this test is to demonstrate the ability of the PORV to open. The probability of a malfunc-tion of equipment important to safety will not be increased by this test. This test is being per-formed to demonstrate the ability of the PORV to open. The consequences of a malfunction of equip-ment important to safety will not be increased by this test. If the PORV were to malfunction, the control room operators can close the block valve, h' In addition, an open PORV or safety valve is ana-lyzed in Section 15.1.4. The margin of safety as defined in the bases to Tech Specs will not be !
reduced by this test. Prior to opening the PORV, secondary thermal power will be verified by proce-dure to be less than or equal to 97% due to slight increase (< 3%) that will occur when the PORV is opened. A temporary test transmitter is installed on the positioner output for testing. This trans- !
mitter measures the differential pressure across the valvc actuator. This transmitter will not adversely affect the ability of the valve to operate. The 1 43
4 transmitter is independently verified to be removed and the positioner returned to normal prior to returning the valve to operable status.
OP/2/A/6350/02 Diesel Generator Operation: FSAR Section 8.3.1.1.3.4 describes the overspeed protective function of tha diesel generators as a pneumatic trip system. NSM CN-11104 and 20486 removed the pneumatic emergency trip system and replaced it with an electronic system. The pneumatic system was unreliable and resulted in several forced outages.
The electronic emergency trips provide the same protection at the same setpoints as the pneumatic system. The word " pneumatic" mentioned in the FSAR with reference to the emergency trips _for the diesel generators needs to be changed to " electronic".
Based on the fact that no change has been made to the function of the emergency trips and according to Design Engineering, better reliability of the pro-tective system is expected. I find no evidence to suggest that an unreviewed safety question exists.
Attached is the 10CFR50.59 evaluation for the implementation of the modification. It reinforces my review.
PT/1(2)/A/4550/03L Post-Refueling Core Loading Verification: This procedure is used to verify that the core has been reloaded with the assemblies in the configuration specified by the Special Nuclear Material (SNM) file core maps. Each core location is checked against the SNM maps to verify correct assembly identifica-tion and orientation and the presence or absence of an RCCA. Per the FSAR safety analysis (Section 15.4.7), this is to be done prior to each restart.
The exact method of performing the verification is not specified in the FSAR nor the applicable APM, Reg Guide, NVREG or standard. The procedure's method involves first recording the assembly iden-tification that and then independently verifying the results against the SNM maps. This method provides good assurance that any misloaded fuel will be detected. This is the standard method used at the other Duke units.
As mentioned in the safety evaluation for PT/1(2)/A/4150/22, Total Core Reloading, there is a
" case" which the FSAR dens not consider. This is the case in which an RCCA is misplaced.
PT/1(2)/A/4150/18, Fuel Assembly - Insert Shuffle Procedure, provides the administrative control over 143
the insert " shuffle" movements in order to preclude this problem. pT/l(2)/A/4550/09, Fuel Assembly -
Insert Verification, provides a check that the RCCAs are in their designated fuel assembly for the next cycle, pT/1(2)/A/4550/03C, Post-Refueling Core Loading Verification, air.o provides a check to verify that RCCAs are located only in core locations l' designated for them.
Additional safety considerations concern the fact that the camera movements directed by this PT are classified as core alterations. This PT is per-formed under the direction of P1/1(2)/A/4150/22, L Total Core Reloading. That PT contains the admin-istrative controls that verify the satisfying of the various core alteration Tech Spec requirements.
These requirements include containment isolation, source range alarms, manpower, minimum boron con-centration and level. The camera, its cable, and any lights weigh less than 100 lbs total and there-fore are within the bounds of the analysis in FSAR Section 15.7.4 which covers fuel handling accidents.
There it is assumed that an entire assembly is dropped. The provision to allow suspension of ND 1 loop operation for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> each 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is permitted by Tech Spec 3.9.8.1. It requires that full level in the cavity be maintained ano that at least one loop of the ND system be operable. Thus, adequate cooling capability is maintained.
Since the methodology of this procedure is within the assumptions made in the FSAR and provides additional protection against RCCA misplacement, the possibility or consequences of an accident previ-ously evaluated in the FSAR will not be increased nor will the possibility of an accident not already evaluated in the FSAR be created. The only equip-ment that could possibly be affected are the RCCAs.
Since this procedure does not physically affect them I but provides additional assurance that they are not 1
- misplaced, the probability or consequences of previ- 1 ously evaluated equipment important to safety will l- not be increased nor will the possibility of mal- i function of equipment important to safety which is 1 different than that already analyzed be created. !
This procedure is performed under th direction of PT/l(2)/A/4150/22, Total Core Reloading. That ,
procedure e;.sures compliance with th various Tech Spec requirements concerning fuel handling and core alterations. No margin of safety as defined in the bases will be reduced, l
144 l
i
r ]
y PT/1/A/4200/55, CA Check Valve Leak Rate Test: This change will Change #3 ensure that the suction piping to the CA Pumps is >
sufficiently filled and vented after performance of this test. The piping is practically drained as a prerequisite to this procedure.
The probability of and consequences of an accident previously evaluated in the FSAR are not increased 1 as a result of this change. The mode in which this
? procedure is performed is not one in which the CA System is required. No new accident scenarios are i created as a result of this change.
The probability of and consequences of equipment important tc safety malfunctions previously evalu-ated in the FSAR are actually decreased as a result '
of this change. The CA pumps will have less of a chance of cavitation or air entrainment since this change ensures that the suction piping is filled and vented after performance of this test. Also, since the piping is being verified water solid, the ;
possibility of e 1 functions of any equipment impor- '
tant to safety not evaluated in the FSAR is not created.
Since this test is performed in a mode that the CA System is not required and this change does not affect the mode, the margin of safety as definea in the bases of the Technical Specifications is not reduced.
OP/1/A/6700/01, Unit One Data Book: OP/1/A/6700/01, Unit One Data Change #153 Book, Table 2.2 is used tr *ecord the 100% Full Power calibration current, (at Axial Offsets of -
+20%, 0% and -20%) and the M Factors for each'of the ,
Power Range Excore Detectors. Data is obtained for this table only by approved procedure, PT/1/A/4600/05A, Incore/Excore calibration. <
The data recorded on this table is used by I&E to adjust the ACD calculating circuitry and 0AC pro-grams. It may also be used to manually calculate AFD and QFTR if the OAC is inoperable.
Since AFD is used to dynamically adjust both the OTAT and OPAT setpoints, the data her is safety related. The FSAR accidents references in this evaluation depend upon these setpoints for their mitigation.
This change will not increase the probability or consequences of any accident either analyzed or unanalyzed by the FSAR. No safety significant equipinent malfunction either analyzed or unanalyzed by the FSAR will be created (no equipment other than i 145
L P
the Power Range NIS is affected by this change).
The margin of safety defined in the bases of the i- affected Tech Specs will not be reduced in any way by this change. ,
PT/2/A/4200/55 CA Check Valve Leak Rate Test: The purpose of this procedure is to measure leakage through check valves 2CA8, 2CA10, and 2CA12. Those check valves provide the boundary to prevent safety grade water from the RN System spilling through a possible pipe break ,
upstream of the check valves and not be available to !
- the CA pumps. y
, This procedure will be performed in either Mode 4, -
! 5, 6, or possibly no mode. The CA System is re-quired for Modes 1, 2, and 3 only, Since the CA System is not required when this procedure is being performed, the probability of and consequences of an accident previcusly evaluated in we FSAR will not be increased. By only making the CA System un-- ;
available, the possibility of an accident different -
from any previously evaluated in the FSAR is nct created.
Prior to performing the valve lineup for the' test, the breakers for the pump motors will be racked out and the isolation valves for the steam to the CA pump turbine will be closed. The valve lineup will isolate suction to the pumps, but there is no enance that the pumps will start. Therefore, the possibility of and consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Since the pumps cannot be started without a suction source and all valves are operated in a normal manner, no new equipment malfunction possibilities are created. .
Since the CA System is not required at the times this procedure will be performed, the margin of ,
, safety as defined in the Technical 9 ecification bases cannot be reduced.
PT/2/A/4350/02E CA, CF, and Turbine Interlocks Periodic Test: The CA, CF, and Turbine Interlocks Periodic Test is diverse in the variety of surveillances it covers.
Therefore, this safety evaluation will be broken into fcur parts.
A. Sections 12.1.1, 12.1.2, and 12.2 Motor-Driven CA Pumps Auto-Starts In this section, each pump is aligned in the recirculation mode separately and response time tested for various automatic start signals.
146
e
~
t Actuation of associated flow control valves is also tested in response to these signals. ,
Accident flowrate is set up in advance. These sections are tested in either Mode 4, 5, 6 or No Mode, when the CA System is not required to be operable.
Therefore, the probability of and consequences of cny accioent already analyzed in the FSAR i are not increased. The possibility of any ;
accident not previously evaluated in the FSAR is not created. Since the pumps are operated-will within mechanical design limits, the C probability of and consequences of any equip- i ment malfunction is not increased. Neither
, will the possibility of any equipment malfunc-l tion be created that would have an impact on nuclear safety. ,
Since this test is performed in modes that the
. CA System is not required, the margin of safety e as defined in the Technical Specification Bases '
cannot be reduced.
B. Sections 12.1.3, 12.1.1, and 12.3 Turbine _ Driven CA Pumpj_AMo-Starts ,
In this section, CA Pump #2 is aligned in the recirculation mode and response time tested for various automatic start signals. Actuation of associated flow control valves is also tested '
in response to these signals. Accident flow rate is set up in advanco. This test must be performed in Modes 1, 2, or 3 since Main Steam !
must be available to run the pump. However, at r least one motor-driven CA Pump will still be operable and both will be available during this testing.
Since this pump will be run in recirculation mode and started by placement of a jumper that effects no other' components, the p ebability of and consequences of any previous 13 evaluated accident is not increased. No accident not analyzed in the FSAR is created. Since the pump will not ce operated outside of design ,
parameters, no equipment malfunction possibil- '
ities or consequences are increased. No new.
equipment malfunction possibilities are created due to this test.
Since the required number of CA Pue s are still available, the margin of safety as defined in 14.1
l the bases of Technical Specifications is not-reduced.
C. Sections 12.4 and 12.7 - CF Isolation These sections will test CF Isolation on Hi-Hi Doghouse Level and Hi-Hi Steam Generator' level.
Doghouse level will be simulated high by placement of jumpers. Steam Generator level
( will be simulated high by manipulation,of -
Process Control Cabinet logic cards and placing one train of SSPS in test. This testing will be performed in Modes 4, 5, 6 or No Mode.
Placing the jumpers for Hi-Hi Doghouse Level does not make any component inoperable, re-gardless of Mode. Process Control Cabinet modifications and placement of one train of SSPS in test are no problem in Modes 5 or 6. :
In Mode 4, however, one train of SSPS will be inoperable due to being in test. The opposite train will, however, still be operable. In addition, manipulation of Process Control Cabinet logic cards will reverse the logic of the Steam Generator Level interlocks if levels are initially in the Hi-Hi state. This will result in all 4 channels of S/G 1evel being failed low. The reactor will already be tripped, and the CA System is not required in Mode 4. Therefore, the probability and conse-quences of any accident previously evaluated in the FSAR will not be increased, and no new ,
accident scenarios are created. In addition, since the only equipment operation is the normal stroking of valves, no equipment mM-function possibilities or consequences are increased. No new equipment malfunction possibilities are created.
Since the valves are being failed to the safety position and only one-train of SSPS is inoper-able at any tiae, the margin of safety as defined in the Technical Specification Bases is not reduced.
D. Sections 12.5. 12.6._12.8. and 12.9 - Turbine -
T r_1 p, I
The Main Turbine will be tripped on the fol-lowing signals: trip of both CF Pumps, Reactor Trip, Hi-Hi S/G Level, and manual trip signal.
The turbine is not safety related and since these trips will be performed with no steam l passing through the turbine, there will be no 148
b l i,
b transient affect on the NC System and no interaction with any safety-related system. ,
Therefore, no safety questions are involved ;
with these sections.
PT/2/A/4200/010 Containment Isolation Valve Leak Rate Test: This :
change will allow testing of check valve 2NM425 with a plug installed in valve 2NM25A. The plug will act as a boundary to ensure all measured leakage is through check valve 2NM425. All leakage measured will be attributed to check valve 2NM425 for acceptance criteria verification.
The plug in 2NM25A is installed.per a work request and will be removed per the work request.- This procedures ensures that the plug is installed before testing the check valve. Performance of the proce- I dure with this umporary change will put increase the consequences or probability of an accident.
OP/2/A/6700/03 Operation With the Operator Aid Computer Out of Service: OP/2/A/6700/03 (Operation With the Opera- ,
tor Aid Computer Out of Service) retype 2 incorpo- !
o rates changes 0 through 4. Additional changes made include:
- Better defining the time requirements for performance of PT/2/A/4600/08B (Manuel Calcu-lation of Quadrant Tilt) and PT/0/A/4220/01 '
(Hand Heat Balance and NC Flow Calculation).
- The addition of a step to monitor the Main Feedwater Isolation Valve Actuator nitrogen press. This ensures operaoility per Tech Spec 3/4.6.3. This' step was added in response to )
JC0 that was written as part of the answer to PIR-0-C88-0099.
This retype does not increase the probability of an - '
accident or equipment malfunction already evaluated in the FSAR, nor does it create a new scenario for these to occur. The consequences of an accident.or equipment malfunction will be unchanged by use of i this procedure. The margin of safety as defined.in i the Tech Specs is unaffected. No :wreviewed safety l question is judged to exist by this retype or its l use. l PT/0/B/4600/14 Chemistry Periodic Surveillance-Items: This Revision 7 revision to PT/0/B/4600/14 is necessary to incorpo-rate new surveillance activities required for placing the CNS Monitor Taak Building (MTB) into !
service. The MTB was constructed under modification number CN-50180 and, as such, has undergone i i,
149 !
- l Lm
t l
p
?'
i extensive safety review via the normal design and construction phases. No further review of the safety or accident analysis is required.
4 The description of the MTB does not yet appear in the FSAR; however, the surveillance requirements were issued by the USNRC as Amendments 38 and 45 to Facility Operating license NPF-35 and NPF-52.
These increased surveillances resulted in the need to revir,e this procedure. This change has no effect on the safety analysis or the margin of safety as defined in Tech Specs.
IP/2/B/3222/56A Process. Control System Steam Generator Level 56B Control: See 50.59 evaluction for NSM CN-20535.
56C TT/2/A/9200/51 ATWS Functional Test for NSM CN-20346: The purpose of th',s test is to functionally verify the instal-lation of NSM CN-20346. This NSM installs the Anticipated Transient Without Scram (ATWS) Mitiga-tion System and Actuation Circuitry (AMSAC). 1he AMSAC will initiate upon the trip of both CFPT's or loss of any 3/4 CF.flowpaths. This procedure will test the following:
- 1) Trip of Main Turbine and start of CA Motor-Driven Pumps upon AMSAC actuation.
- 2) Actuation of AMSAC upon trip of both CFPT's.
flowpaths.
- 4) Block of AMSAC actuation by both manual and -
automatic inputs.
- 5) Verification of various CF flowpath blockage inputs.
The AMSAC is not required by Technical Specifica-tions, nor is it a part of any FSAR accident analy-sis. Therefore, the probability of and consequences of any accident previously evaluated in the FSAR are not increased. Since the test is run in Modes 5 or 6 when neither containment isolation, feedwater isolation, nor auxiliary feedwater are required, no ,
new accident scenarios are created.
All valves in this procedure are stroked in a normal fashion. The CA Pumps are not actually run, as the breakers remain in the " test" position throughout the procedure. Therefore, the probability of and 150
.s h-p consequences of a malfunction of equipment important to safety previously analyzed in the FSAR are not increased. Since no abnormal operation of equipment is performed, no new malfunctions of equipment important to safety are created.
The margin of safety as defined in Tech Specs is not reduced for two reasons: 1) the AMSAC is not
, addressed in Tech Specs and 2) none of the systems affected in this test are required in Modes 5 or 6.
I PT/1/A/4350/02E, CA, CF, and Turbine Interlocks Periodic Test: This Change #42 promdure change is being written to allow retesting-o f v. .' logic to valves 1SA2 and ISAS without actu-ally running the CA Turbine-Driven Pump. Steps requiring starting and response-time testing of the pump are deleted, and block valves ISA1 and ISA4 are added to the valve lineup and are verified closed. -
Since steam to the CAPT will be isolated during this test, the CApl will be inoperable. Hcwever, the 10CFR50.59 analysis for the procedure (dated 6/28/88) has already addressed CA pump operability (since the procedure normally aligns the CAPT to run in recirculation, which also makes the pur p inoper-able).
Therefore, the probability / consequences of any new or previously-analyzed accident or equipment mal-function is not increased, and the margin of safety is not reduced.
PT/1/A/4250/060, CA System Flow Verification: The purpose of this Change #5 change is to replace the enclosure for CA Pump #1 flow verification. The maximum pump speed has been increased from 3620 RPM to 3650, resulting in higher pump performance. The new enclosure reflects values for flow balancing provided by Design Engineering per CNDS-1223.42-0011, Rev. 3.
The procedure change will not increase the proba-bility of or consequences of any accident previously evaluated in the FSAR. The new values have been calculated and verified by Design Engineering for acceptable CA flows for postulated accidents. No new accident scenarios are created by this change since the test method is not revised.
The possibility of and consequences of equipment malfunctions are not increased. The flowrates and speed are well within the design limitations of the pump. This change does not affect the manner in which the pump is operated.
1_51
i k !
l 4 This change does not make any of the CA System.
inoperab'e, nor does it lessen the ability of the
, system to mitigate the consequences of an accident.
L -
Therefore, the margin of safety as defined in the
! bases to the Technict.i Specifications is not re- :
duced.
7 OP/1/A/6700/01 Unit One Data Book: The Target AFD in Data Book '
L Figure 1.1 is updated by performance of L
PT/1/A/4150/08, Target Flux Difference Calculation.
The target AFD is changed to keep Control Bank D at approximately 215 steps withdrawn (at 100% Full-Power) while allowing for ch:nges in the natural axial power shape that occur with burnup. The ,
targets are set by procedure to be within the operating bounds of T.S. 3/4.2.1. The targets are an operating guideline only to aid the control room
, operators in maintaining AFD within the limits of ,
'T.S. 3/4.2.1. The targets serve no other purpose. .
They do not feed any trip function or serve any . '
safety related functions. The limits that must be observed Tn T.S. 3/4.2.1 are see by cycle specific analysis.
AFD is an input to the OTDT Trip Setpoint. Targets may be set so that the AFD input function to the '
OTDT will impose a penalty on OTDT. However, this penalty will automatically 50 imposed by the.7300 system per the formulation of T.S. 2.2.1. -
AFD and AFD targets are monitored by the OAC NUCLEAR 06 (AFD ALARM HONITOR) program.. AFD Targets on the l OAC are changed by the above mentioned procedure, ,
also. (
All accidents analyzed in FSAR Chapter 15 have as one of the initial conditions that AFD is within the limits of T.S. 3/4.2.1. As such, the targets do not have any effect on the accident analysis.
OP/2/A/6350/02 Dies'l Generator Operation: FSAR Section 8.3,1,1.3.4 describes the overspeed protective function of the diesel generators as a pneumatic trip system. NSM CN-11104 and 20486 removed the pneumatic emergencj trip system and replaced it with an electronic system. The pneumatic system was unireliable and resulted in several forced outages.
The electronic emergency trips provide the same protection at the same setpoints as the pneumatic system. The word " pneumatic" mentioned in the FSAR with reference to the emergency trips for the diesel generators needs to be changed to " electronic".
1 51
r
-l r
[,3 Based on the fact that no change has been made to the function of the emergency trips and according to Design Engineering better reliability of the pro-tective system is expected I find no evidence to L suggest that an unreviewed safety question exists.
OP/2/A/6100/09A Annotator Response W ?A D/G Panel: ~FSAR Section E 8.3.1.1.3.4 descril m 0 o overspeed protective function of the diem .;<nerators as a pneumatic trip system. NSM CN-11104 and 20486 removed the pneumatic emergency trip system and replaced it with an electronic system. The pneumatic system was -
unreliable and resulted in several forced outages.
. The electronic emergency trips provide the same protection at the same setpoints as the pneumatic 1 . system. The word " pneumatic" mentioned in the FSAR with reference to the emergency trips for the diesel generators needs to be changed to " electronic".
Based on the fact that no change has been made to the function of the emergency trips and tecording to Design Engineering, better reliability of the pro-tective system.is expected, l. find no evidence to suggest that an unreviewed safety question exists.
MP/0/A/7600/62 Fisher Pressure Regulating Valve Corrective Mainte-nance: .This procedure performs corrective mainte-nance'on Fisher Pressure Regulators. Instruction manual CNM 1201.05-0255 provided technical informa-tion in the development of the procedure. Plant specific information that applin to Catawba Nuclear-Station was also included in.its development.
This evaluation is for changes made during the procedure upgrade process. These changes are in the procedure format and are not significant-in content.
The Catawba FSAR and Technical Specifications 'have been reviewed and are not affected by this proce-dure. This procedure applies to valves used in various plant applications; This procedure will be used.to correct and improve the performance of the valve and will maintain the valve within its-origi--
nal design requirements and specifications. -The safety standards previously specified-within the i FSAR and' Technical Specifications will be main-tained. The probability of an accident or malfunc-tion previonly addressed in the FSAR will not be increased. Ho unreviewed safety qu?stions are involved.
MP/0/A/7150/39 Reactor Coulant Pump Seal Removal and Replacement:
- This safety evaluation is for the revision of MP/0/A/7150/39 changes 0 to 12 incorporated, 153 s
' 'Tg;
y L
L i
L prepared on 04-10-89. This procedure has been revised to incorporate lessens learned and correc-tions identified during procedure use. These :
revisions and changes will increase the reliability ,
of the work controlled by this procedure. The changes to this procedure are identified on the attached marked up pages of the old procedure.
Tech Specs 3.2.3, 3.4.1, and 5.4 are affected by this procedure. Operationr, has the responsibility
- and the procedures for compliance with these Tech
[ Specs. Maintenance will be performed on this pump when Tech Specs allow, per Operations' procedures.
These revisions will clarify and assure that main-F tenance activities will return the Reactor-Coolant Pump Seals to as-designed conditions.
A full USQ evaluation is required because this procedure is being changed significainly and it is described in the FSAR. FSAR Section 13.5.2.2.1 (Maintenance Procedures) states that i,.aintenance procedures are required and since this procedure thange, as described above, changes this procedure in a significant manner, a USQ evaluation must be performed.
The corrections made by this procedure change have been reviewed against approved vendor manuals, j design documents, and station procedures to ensure 1 that-the corrective maintenance controlled by this '
procedure will return th pump to as-built /as-designed condition. These actions will ensure the pump's compliance with FSAR accident analysis. Since the pump will be returned to as-designed conditions, the possibility, consequences, or probability of a malfunction will be reduced. The possibility, probability or, consequences of a previously unre-viewed safety question are not created by this change because the pump will be returned to as-designed conditions. By the same reasoning, the margin of safety as defined in the Bases of Tech Specs will i not be reduced. Therefore, no 35Q exists.
PT/0/A/4400/06 Essential Heat Exchangers Heat Capacity Test: This procedure change will allow throttling of the RN flow to KD Heat Exchanger 2B. Because of this D/G 20 will be declared inoperable due to RN flow balance criteria not being met. The lower RN flow will give better results due to better stability of the KD System. No USQ was deemed to exist.
OP/2/A/6550/07 Reactor Building Manipulator Crane Operation: 1 Op/2/A/6550/07 retype 2 reflects changes 1 and 2 l
1 53
y, +
s along with additional changes incorporated into OP/1/A/6550/07 retype 2. Unit 2 ZZ tape readings are correct as is. In retrospect, Step 2.18 of Enclosure 4.1 was expanded upon to provide more instruction on obtaining " ROD GRIFPER FULL UP position. Steps 2.11 and 2.18 of Enclosure 4.1 makes reference to a new additional Enc 1csure (4.13) " Raising and Lowering of Rod Inner-mast / Gripper using By-Pass Switches". Enclosure 4.13 was added to the procedure after the vandor, Sterns and Rodger's explained the importance of raising the Rod gripper to the ROD GRIPPER FULL VP position. This position lessens the tension placed on the various reels; Example: ZZ tape reels, and in essence keeps the reels in good working order.
This enclosure provides the operator with detailed directions on rsising and loweririg the Rod mast.- If operations other than core alterations are to be performed, the Rod mast latches may be placed in bypass to perform crane movements. By using the latch bypass with the crane in this configuration, the need for bridge and trolley bypasses is elimi-nated, if moving the crane within its normal oper-ating area. The Latch Sol bypass places the Latches in a position to provide the required interlock to allow bridge and trolley movement.
Paragraph 9.1.4.3.1(6) of the FSAR references bridge and trolley movements except when the fuel gripper is raised to the " Fuel Gripper Up disengaged" posi-tion or when the Rod Gripper is raised to the " Rod Gripper Up First Stage" position.
By operating the crane with the Rod Gripper in the full up position, with the latches bypassed to {
retract position, the intent of this paragraph is met. There is no increase in the chance of damaging the equipment by operating in this fashion, When Core Alterations are to take place, the crane is returned to the configuration described in the FSAR.
Numerous editorial changes have been made throughout' i the procedure. (
Op/1/A/6550/07 Reactor Building Manipulator Crane Operation: i OP/1/A/6550/07 retype 2 incorporates changes 1 and 2 A along with additional changes. Upon further obser-vation and or use of the reactor manipulator crane, it was determined that the ZZ tape readings for R00 and TUBE as per Enclosure 4.7 needed to be trans-posed. In u trospect, Step 2.18 of Enclosure 4.1 was expando.1 upon to provide more instruction on ob- ,
taining " ROD GRIPPER FULL VP" position. Steps 2.11 and 2.18 of Enclosure 4.1 makes reference to a nee j 155 T
k L
l additional Enclosure (4.13)" Raising and Lowering of
- Rod Innermast/ Gripper using By pass Switches".
Enclosure 4.13 was added to the procedure after the vendor, Sterns and Rodgers explained the importance of raising the Rod gripper to the ROD GRIPPER FULL UP position. This position lessens the tension placed on the various reels, Example: ZZ tape reels, and in essence keeps the reels in good working order, t
[ This enclosure provides the operator with detailed l directions on raising and lowering the Rod mast. If operations other than core alterations are to be e
performed, the Rod mast latches may be placed in l "- bypass to perform crane movements. By using the latch bypass with the crane in this configuration, the need for bridge and trolley bypasses is elimi-o nated if moving the crane within its normal operat-ing area. The Latch Sol bypass places the latches in a position to provide the required interlock to allow bridge and trolley movement.
Paragraph 9.1.4.3.1(6) of the FSAR references bridge and trolley movements except when the fuel gripper is raised to the " Fuel Gripper Up Disengaged" post-tion or when the Rod Gripper is raised to the " Rod Gripper Up First Stage" position.
By operating the crane with the Rod Gripper in the full up position, with the latches bypassed to retract position, the intent of this paragraph is met. There is no increase in the chance of damaging the equipment by operating in this fashion. When Core Alterations are to take place the crane is returned to the configuration described in the FSAR.
Numercus editorial changes have been made throughout ,
the procedure.
')
Op/1/A/6550/06 & Transferring Fuel With the Spent Fuel Manipulator P
OP/2/A/6550/06 Crane: OP/2/A/6550/06 retype 4 Enclosure 4.1, which previously was titled, Startup of the Spent Fuel Pool Manipulator Crane, and Enclosure 4.3 previously titled Securing the Spent Fuel Pool Manipulator-Crane have been incorporated into one Enclosure, titled 4.1 Startup and Shutdown of the Spent Fuel Pool Manipulator Crane. Note: Securing was changed to Shutdown. The phrase Shutdown is used in the Reactor-Startup and Shutdown enclosure. In turn, the most frequently obtained enclosures in order of use, is the Startup and Shutdown enclosure followed by 4.9 Temporary Shatdown and Startup of the Spent ;
Fuel Manipulator Crane. This enclosure has been !
changed from 4.0 to 4.2. Steps have been added to ,
enclosures 4.1 and 4.2 to instruct the operator !
i
1 t,
depending on whether he or she is starting the crane or shutting it down, to lower the Fuel Hoist from a Gripper Tube Up position to Gripper Up Disengar d ~1 position during startup and when shutting the trane ;
down for extended periods of time, to place the
-Gripper Up Disengaged By-Pass switch to the "0N" L position and raise the Fuel Hoist to Gripper Tube Up 1 position. These additional steps evolved after the b crane vendor, " Sterns and Rodgers" explained that !
h these measures would lessen the. tension placed on j the various reels, Example: ZZ tape. l
! Step 2.5 Relock the control console of Enclosure 4.1 ,
[
~
was deleted, but deferred to the shutdown part of i Enclosure 4.1 and reinstated as 2.16 Lock the crane console. The original step was premature. ;
i p The inverse opposite was changed on Enclosure 4.2 ,
- Temporary Shutdown and Startup of Spent Fuel Manip- .
L- ulator Crane, but with exception to Relocking; this 1 step is concluded once again in 4.1 enclosure i t
(Shutdown Section).
Step 1.2 of Enclosure 4.2 now 4.3 Transferring Fuel j has been deleted and inserted as Step 1.4 of Enclo-- 1 L
sure 4.1 Startup and Shutdown procedure. The crane !
is required for transferring fuel and it is more
. appropriate to include this OPS signoff with the others. Permanent tags with the following ZZ tcpe
, readings; Gripper. Up Disengaged, Gripper Tube Up, Upender, New Fuel Elevator, Load Test Stand will be posted prior to implementation of the following statement. .
t' L. -Quote: or per posted "ZZ" - tape readings on the crane console. The tag will be attached to the t crane. Enclosure 4.4 Fuel Assembly and Inserts j Weights and Associated Drag Forces, contains weights for assemblies etc. that are handled by the crane >
and the fuel handling tools combined. This a enclosure has been deleted and replaced with~a more
- lr concise enclosure tabulated from the procedure r OP/2/A/6550/07 Reactor Woights, and pertains to movements with the crane only. This ent- wre was i
[ used for the new 4.4. .
In Enclosure 4.7 a setpoint calculation was added ,
for the New Fuel Elevator (NFE) because no design information is available for setpoints. The calcu-lations for the three setpoints are based on the following.
D
< i l
1 No Load: The purpose of this interlock is to stop j
. downward movement of the elevator when load on the I hoist cable is lost. This calculation of 1/2 the i NFE weight down is an engineering judgement. By using this valve the elevator will be allowed to travel down with no additional weight on it, but it will stop, if for some reason the weight on the i cables was lost prior to the cables going completely slack. ;
Normal Load: The purpose of this interlock is.to I disallow the elevator to be reised with an assembly p in it without the use of the bypass circuit. The '
E calculation of NFE weight up to 1/4 the weight of a ,
L wet assembly is an engineering judgement. By using i L this valve the elevator will be allowed to come up 4 empty with normal mechanical vibrations not triaping P the circuit out. If an assembly is placed in tie NFE, it will not raise the assembly without the use of the Bypass. ;
Overload: The purpose of this interlock is to stop i the NFE when raising it, and with the Dummy Fuel assembly as cargo and the Bypass switch "0N". This i interlock will also stop the NFE going down if it is !
overloaded. -!
The calculations of NFE weight up to weight of a' dry assembly and rod 4200 lbs ensures the capability to "
raise the assembly using the Bypass, but does not overload the NFE hoist if the NFE should become= '
bound for some reason, The additional-200 lbs 1s:
added to allow for normal vibrations in the system !
as it moves. These vibrations become translated -;
into weight seen by the load cell.
The use of this procedure does not change the-probability or consequences of any accident or i equipment malfunction discussed in the FSAR. Nor .)
does it' create the possibility of a.different -l' accident or equipment malfunction scenario. The margin of safety as defined in the Bases for Tech s Specs is' unaffected by this procedure. This proce-dure ensures Tech Spec requirements for crane and fuel movements are being met. No unreviewed safety- s
( questions are judged to exist by this procedure or its use. ;
IP/1/A/3110/07 Calibration Procedure for Component Cooling System l IP/2/A/3110/07 Hert Exchanger A & B: See 50.59 evaluation for exempt enange VN CE-2129.
PT/1/A/4550/010 & Reactor Building Manipulator Crane Load Test:
L PT/2/A/4550/010- PT/1/A/4550/01D retype #1 changes were made to reflect a more accurate identification of the i 158 ;
l
- y. ,
l equipment used in the performance of this procedure. 1 These changes were made to show for example an )
indication light, switch (by-pass) or controls as i they are actually ID tagged to the person (s) per- 3 forming the procedure. Additional information on valve arid control locations was included in partic-ular stops. A number of steps were in incorrect order ar.d were rearranged so an action performed _
corresponded with a result achieved.. Furthermore, a i number of enclosures requir'ed addi.tional steps within the sequence to emphasize actual order of
-events oa equipment responses. In addition to the-above, changes 1-3 are incorporated in this proce- >
dure.
The use cf this procedure does not change the i probability or consequences of any' accident or en
- ment malfunction discussed in the FSAR. Nor oe s it create the possibility of a different :
accident >r equipment malfunction scenaric. The. ';
margin of safety as defined in the bases for Tech i Spec 3/4.9.6 is unaffected by use of.this procedure, :
because this procedure ensures the surveillance- ,
requirements for this Tech Spec are met. >
OP/1/A/6700/03 Operation With the Operator Aid Computer Out of -
Retype #4 Service: OP/1/A/6700/03 (Operation W1th- the Opera- -
tor Aid Computer Out of Service) retype #4 incorpo- ;
rates changes 4 through 7. Additional changes made include:
- Better defining the time requirements for performance of PT/1/A/4600/088 (Manual Calcu-lation of Quadrant Tilt) and.PT/0/A/4220/01 .
(lland Heat Balance and NC Flow Calculation).
- The addition of a step to monitor the Main.
Feedwater Isolation Valve Actuator nitrogen ;
pressure. This ensures operability per Tech )
Spec 3/4.6.3. This' step was added in response o to JC0 that was written as part of-the ANS'to i pIR-0-088-0099. !
- Deleted requirement for Tech Spec Logbook entry -;'
, for Pressurizer PORV indication. A change to Tech Specs deletes this requirement of using OAC for this indication. ,
This retype does not increase the probability of an accident or equipment malfunction already evaluated in the FSAR, nor does it create a new scenario for these to occur. The consequences of an accident or- :
equipment malfunction will be unchanged by use~of this procedure. The margin of safety as defined in ee .
, ~~.
g 3 b
4 the Tech Specs is unaffected. No unreviewed safety question is judged to exist by this retype or its use.
"P/2/A/6550/14
.i Draining and Filling of Spent Fuel Pool Transfer Retype #0 Canal and Cask Area: OP/2/A/6550/14 (Drain and Filling of the Spent Fuel Pool Transfer Canal and Cask Area) retype #0 was originally OP/0/A/6550/14.
This new procedure has been developed to split the units as required by OMP 4-1. Additional changes have been included, most to provide the user with additional guidance and to ease the use of the procedure. These changes include:
Limits and Precautions on the possibility of-reducing boron concentration in the Spent Fuel Pool when washing down the walls with domin water.
Limit and Precaution to maintain either 2KF122 closed or the blind flange in place while the weir gate is in place. This is in response to IE 88-92.
Limit and Precaution to inform the operator not to inflate the weir gate seals unless they are installed. 'The manufacturer will allow the gates to be inflated to 15 psig but there is no need to inflate them unless installed.
Additional steps were added to provide adequate protection of the Standby Makeup Pump tagging the pump out and better seal pressure monitor-ing directions were provided.
Numerous editorial revisions were made throughout to-improve the procedure's usability. No.
operational change was made by them.
The use of this procedure will not increase the probability or consequences of any accident already evaluated in the FSAR. Nor will it create a new accident scenario. The probability or consequences of a malfunction of equipment important to safety is j unaffected by the use of this procedure. No new '
scenario that could cause a malfunction of equipment-is created by the use of this procedure. The margin of safety as defined for Tech Specs 3/4.9.7 and 3/4.9.13 will be assured to be met by use of this procedure. No unreviewed safety question is judged to exist by this procedure or its use.
l 160 l
F l
, PT/2/A/4700/01 Periodic Test Performance Verification: Previously the interval period was "9 days" on one pump every 27 days and this allowed the test to fall on a weekend from time to time. To prevent this, the Unit Coordinators changed the scheduling of the PT so it will be performed on a weekday so essential personnel are available if problems occur. The 28 day interval period remains within the Tech Spec Surveillance frequency of 4.7.1.2.1 which is once per 31 days. This change provides consistency between the Interval period listed in the PT and the actual performance of PT/2/A/4250/06. This change will not degrade the operation of a safety system during a design basis accident.
PT/1/A/4200/26 NS Valve Inservice Test (QU): The change in ques-tion removes 3 valves from the test alignment.
These 3 valves were aligned to prevent spraying water into containment through the spray header when INS-32A is opened. Currently the NS System is tagged out and drained so these 3 valves are not-required to be aligned. Therefore, all questions above are answered "No".
'PT/1/A/4450/05A Containment Air Return Fan and Hydrogen Skimmer Fan 1A Performance Test: Steps were added in Sections )
8,0 and 12.1 to remove power from damper ARF-D-2 while the fan (ARF-1A) is operating and return the i power after the fan is shut down. This change '
reducas the possibility of opening ARF-D-2 and inadvertently opening the ice condenser doors during operation of the Air Return Fan. The notes on Enclosures 13.1 and 13.2 were deleted because they are no longer applicable to the procedure. 4 Changes were made to Section 12.3 to operate the Hydrogen Skimmer Fan only momentarily with the suction valve open and to open the suction valve (1VX1A) using the " safety" circuitry. These changes l reduce the possibility of steam bypassing the Ice '{
Condenser in the event of an accident (LOCA) during i the test and use the proper circuitry when opening j 1 IVX1A, Acceptance criteria 11.6 was modified to '
require verification of the start time delay for l HSF-1A. This is a conservative change because the i start time delay is also verified in acceptance criteria 11.7. Jumpers are placed instead of using
" TEST" selector switches and placement / removal of all jumpers is independently verified within the ;
i procedure. Because no " TEST" selector switches are i used, they were deleted from Enclosure 13.3 and I
verified to be in the correct position within Section 12.3. Reference to a 15 minute run time and 161
i b
L ,
l f:
tecording the run time were deleted because accep-tance criteria 11.7 is satisfied under Section 12.4.
1 Changes were made to Section 12.4 to satisfy accep-tance criterias 11.7 and 11.8 within the same section of the procedure. Jumpers are placed instead of using " TEST" selector switches and placement / removal of all jumpers is independently verified within the procedure. Because no " TEST"
, selector switches are used, they were deleted from .
Enclosure 13.4 and verified to be in the correct "
L position within Section 12.4. All reference to the Response Time Testing Program (or a-stopwatch) and >
steps relating to time were added to satisfy .
acceptance criteria 11.7. !
The Containment Air Return Fan IA_(CARF-1A) and Hydrogen Skimmer Fan IA (HSF-1A) are declared inoperable during the performance of this test. The CARF-1B and HSF-1B will remain operable for the duration of the test as required by Tech Spec. None !
of the above changes will prolong the time of j inoperability or place the systems in any unusual alignments that would create or increase the proba-bility of an accident. The margin of safety as defined in the bases of Tech Spec will not be reduced. Placement / removal of jumpers, open-ing/ closing of breakers and returning selector switches to the "As Found" position are indepen-dently verified within Section 12.0 or 13.0 of the-procedure. For these reasons and the ones stated above, these procedure changes do not create or increase the probability of a malfunction of equip-ment important to safety or increase the conse-quences of an accident. Therefore, an unreviewed safety question does not exist. '
IP/1/A/3112/11 Component Cooling Heat Exchanger Outlet Temperature Control: See 50.59 evaluation for exempt change VN CE-1309.
IP/1/A/3030/07H Maintenance and Functional Test Procedure for Main Steam Isolation Valve: See 50.59 evaluation for exempt change VN CE-2258.
IP/0/A/3820/20 Namco IE Limit Switch Installation Requirements:
See 50.59 evaluation for exempt change VN CE-2186.
PT/1/A/4150/07 Administrative Controls for Periodic Testing of Unit 1 Safety Valves: This procedure change identifies the testing completed on Unit 1 safety relief valves per the requirements of ASME Section XI, Subsection IWV-3510. No changes were made that affect plant equipment or operating condition. The FSAR and Tech l
r; ,
t, i
? i
! Spec are not affected. No unreviewed safety ques-tions exist.
PT/1/A/4150/26 Administrative Controls for Periodic Testing of Unit 1 1 Check Valves: This procedure change identifies i the testing completed on Unit 1 for check valves i that were mechanical exercised per th( requirements of ASME Section XI, Subsection IWV-3250. Also added >
- t. is standing work request information of ISA-3 and L 2SA-6, No changes are made that will affect plant equipment or operating conditions.- The FSAR and L> Tech Spec are not affected. No unreviewed safety questions exist. .
OP/2/A/6250/08 Steam Generator Blowdown: This retype includes the >
[ .
Retype #4 following changes: ,
- 1) Previously approved changes 5 through 7 were .;
incorporated.
- 2) locations for manual valves were added in the body of the procedure.
- 3) Numerous clerical changes were incorporated.
- 4) NSM 20363 was completed.
This required numer-ous changes to the procedure to provide guid-ance on operating the BB system with Cold. Water .
Injection permanently installed. ,
- 5) Old Step 2.6 of Enclosure 4.1 to verify open 2BB-188 (BB Pre-Filt to Demin Isolation) was deleted. Thit step was not needed because the valve was opened in the following step per direction of Chemistry _ procedure OP/0/B/6250/15.
- 6) Changes were made to Enclosure 4.4 to ensure applicable Tech Spec entries are made during 2 EMF-34 inoperability.
- 7) Instruction was added to Enclosure 4.4 as to ;
what pushbutton to push to retu,n 2 EMF-34 to >
operation.
This retype will not change the function of the BB System. The changes made will result in a more efficient means of operating the system. The changes included in this retype will only affect the non-safety related portion of the BB System located in the Turbine Building.
- The char.ges included in this retype do not involve or affect any part of a system, including the L> 163 H
[
I e
i containment isolation portion of the BB System, associated with any accident discussed in the FSAR.
The BB System is not an initiator for any accidents l previously evaluated in the FSAR or any new acci-L c'ents; therefore, the probability of their occur-rence will not increase due to the changes made.
The consequences of any accidents will not be increased since the changes included in this retype do not affect any accident mitigating system or reactor coolant system parameters. No equipment important to safety is directly affected by the changes made.
With seismic and overpressure considered,-the probability or consequences of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased. Changes included in this retype will not create the possibility of malfunctions of equipment important to safety different than any already evaluated in the FSAR.
Since no functional changes are being made.to any safety systems and there are no relevant parameters in the bases to any Technical Specification, the margin of safety as defined in the Bases to any Technical Specification is not reduced.
Discussions with J. P. Bolton (MBCE)' informs me that MCNE (Mechanical Catawha Nuclear Engineering) will be revising the FSAR to include operations of the BB System with Cold Water Injection.
MP/0/An600/67 Yarway Y-Type Globe Valve Corrective Maintenance:
This procedure performs corrective maintenance on Yarway Y-Type Globe Valves. Instruction manual CNM-1205.01-0142 provided technical information in the development of the procedure. Plant specific information that applies to Catawba Nuclear Station
, was also included in its development.
This evaluation is for changes made during the l procedure upgrade process. These changes are in the l procedure format and are not significant in content, j The Catawba FSAR and Technical Specifications have '
been reviewed and are not affected by this proce- l dure. This procedure applies to valves used in various plant applications. This procedure will be used to correct and improve the-performance of the valve and will maintain the valve within its origi-nal design requirements and specifications. The safety standards previously specified within the FSAR and Technical Specifications will be main-tained. The probability of an accident or 151
p ,
malfunction previously addressed in the FSAR will not be increased. No unreviewed safety questions are involved.
t
, OP/2/A/6450/10 Operation of the H2 Skimmer fans for Troubleshoot-L ing: Enclosure 4.19 of OP/2/A/6450/10 Containment L Hydrogen Control Systems was written to support the r troubleshooting of the Unit 2 Hydrogen Skimmer Fans.
The procedure was written in such a manner as to L- prevent exceeding motor starting duty for the fans -
Tech Spec motor current while at dead head condi-tions, and lower containment to upper containment
(, air bypass leakage flowpaths.
This procedure also requires that during the trou-r' bleshooting the Hydrogen Skimmer Fan is declared :
inoperable and the Tech Spec Action is-entered. For !
, this situation, the fan may only remain Inoperable for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before shutdown is to commence. During this period, the peak pressure in ,
containment would not be affected due to the opera- ;
bility of the 100% capacity, opposite train, Hydro-gen Skimmer Fan should a LOCA occur and hydrogen :
mitigation be required.
As the operation of the Hydrogen Skimmer Fans 4 dead-headed is a previously evaluated event, no new, !
unreviewed safety questions are created by the use .
of this procedure. Compliance with the recommended {
starting duty, and verification of correct motor amperes assure that no increase in the_ probability or consequences of a previously analyzed accident >
will occur. '
OP/1/A/6450/10 Operation of the H2 Skimmer Fans for Troubleshoot- +
ing: Enclosure 4,19 of OP/1/A/6450/10, Containment ,
Hydrogen Control Systems was written to support the :
troubleshooting of the Unit 1 Hydrogen Skimmer Fans.
The procedure was written in such a manner as to ,
prevent exceeding motor starting duty for the fans, Tech Spec. motor current while at dead head condt-tions, and lower containment to upper containment air bypass leakage flowpaths.
This procedure also requires that during the trou- :
bleshooting the Hydrogen Skimmer Fan is declared 4 inoperable and the Tech Spec Action is entered. For .
this situation, the fan may only remain inoperable i for a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before shutdown is to commence. During this period, the peak pressure in containment would not be affected due to the opera-bility of the 100% capacity, opposite train, Hydro-gen Skimmer Fan should a LOCA occur and hydrogen 't mitigation be required. )
9 As the operation of the Hydrogen Skimmer Fans dead-headed is a previously evaluated event, no new, un.eviewed safety questions are created by the use of this procedurt Compliance with the recommended starting duty, and verification of correct motor amperes assure that no increase in the probability or consequences of a previously analyzed accident will occur.
MP/0/A/7600/28 Yarway 1/2", 1/4", 1", and 2" Piston Check Valve Corrective Maintenance: This procedure performs corrective maintenance on Yarway piston type check valves.
This evaluation is for changes made dur ng the procedure upgrade process. These changes are in the procedure format and are not significant in content.
This procedure and the change will be used to correct and improve the performance of the valve and will maintain the valve within it s original design requirements and specifications. The Catawba FSAR and Technical Specifications have oeen reviewed and are not affected by this procedure. This ' procedure applies to valves used in various plant applica-tions. This procedure will maintain the salves in all applications to the safety sta51dards specified within the Catawba FSAR and Technical 3pecifica-tions. The probability of an accident. on a mal-function previously addressed in t'ne FSAR will not be increased. No new unreviewed safety questions are involved.
PT/1/A/4250/030 RN to CA Pumps Suction Transfer Periodic Test: The purpose of this procedure is to measure the time required for the CA System to switch to the safety grade suction source, the RN System. The motor-driven pumps are isolated by racking out the motor breakers, and the turbine-driven pump is isolated by closing the steam supply manual valves.
The suction piping is aligned so as to introduce as little RN water to the system as possible. The RN to CA suction valves are closed, and a loss of the suction signal is generated. The time required for these valves to go completely open from the time the loss of suctlon signal is genereted is ineasured and ased to verify Surveillance Requirement 4.7.1.2.1.a.4. This test will be performed in Modes 4, S or 6. The CA System is only required in Modes 1, 2, or 3. The RN Sy3 tem operability is not affected by this procedure.
Since ne system is being made inoperable as a result of this test, neither the probability of nor conse-quences of any accident previously evaluated in the 16s
FSAR are increased. No accidents different than those already evaluated in the FSAR are created.
Since the equipment associated with the CA System is not required in Modes 4, 5, or 6, the probability of and consequences of a malfunction of equipment important to safety previously evaluated in the FSAR are not increased. The CA pump breakers will be racked out during this test, and the valves will not be operated abnormally. Therefore, the possibility of malfunctions of equipment important to safety different from those already evaluated in the FSAR is not created.
- ; Since the CA System is not required in Modes 4, 5, or 6, the margin if safety as defined in the Tech-nical Specification Bases are not reduced, iv PT/2/.4 4250/03D RN to CA Pumps Suction Transfer Periodic Test: The purpose of this procedure is to measure the time required for the CA System to switch to the safety grade suction source, the RN Systen. The motor-driven pumps are 1solated by racking out the motor breakers, and the turbine-driven pump is isolated by closing the steam supply manual valves.
The suction piping is aligned so as to introduce as little RN water to the system at possible. The RN to CA suction valvet are clo' sed, and a loss of suction signal is generated. The time requiced for these valves to go completely open from the time the loss of suction signal is generated is measured and used to verify Surveillance Requirement 4.7.1.2.1.a.4. test will be performed in Modes 4, 5, or 6. The CA System is only required in Modes 1, 2, or 3. The RN System operability is not affected by this procedure.
Since no system is being made inoperable es a resuit of this test, neither the probability of nor conse-quences of any accident previously evaluated in the FSAR are increased. No accidents differ?nt than those already evaluated in the FSAR are created.
1 Since the equipment associated with the CA System is not required in Modes 4, 5, or 6, the probability of and consequences of a malfunction af equipment important to safety previously evaluated in the FSAP are not increased. The CA pump breakers will be racked out during this test, and tne velves will not be operated abnormally. Therefore, the possibility of malfunctions of equipment important to safety different from those already evaluated in the FSAR is not created.
167 f.
~
[
Since the CA System is not required.in Modes 4, 5,
.or 6, the margin of safety as defined in'the Tech--
nical Specification Bases are_not reduced.
PT/1/A/4400/06F KD Heat Exchanger IB Heat Capacity Test:- This procedure measures the shell (KD) and tube side (RN) inlet and outlet-temperatures as well as tube side flow in order to determine the shell side fouling -
factor for the KD 1B Heat Exchanger. When RN temperature is low, 1RN-296 needs to be. throttled back to obtain stable test readings. This position .
is different from the RN flow balance position, so the Diesel Generator (D/G)Jis-inoperable. Steps ,
were added to notify Operations that D/G 1B is -
^
inoperable when the position of 1RN-296 is changed and is operable after 1RN-296 has been returned.to 1 the "As'Found" position. The dLOG computer system-is used to acquire data, perform calculations and '
build the data set for the Design Engineering >
program that calculates the shell side fouling factor. This computer system is benchmarked and all_ 1 dLOG programs used by this test and parameters input into to the system are verified correct by the "
i procedure. Measurement of the KD temperature entering the Lube 011 heat exchanger was added for.
information only and does not enter calculations for the fouling factor, a
This procedure is performed while the Diesel Gener-ator is operating in the normal lineup. Shell and tube side temperatures are monitored during test to ensure they stay within design limits. D/G 1B is normally declared inoperable _for less than two hours <
when throttling of 1RN-296 is required and D/G 1A is operable for this period as required by Tech Specs..
So the margin of safety as defined in the basis of 4 Tech Spec will not be reduced. This test does not- 4 involve any jumpers, sliding links, unusual align- .i ments or any other modifications to safety related equipment that would create or increase the proba-bility of an accident. Removing test equipment connected to the heat exchanger and replacing of any-removed process instrumentation are independently verified within the procedure. For these reasons s and the ones stated above, this procedure does not create or increase the probability of a malfunction of equipment important to- safety or increase the ;
consequences of an accident. Therefore, an unreviewed safety question does not exist. -
PT/0/A/4400/22B, Nuclear Service Water Train B Performance Test:
Change #21 The primary purpose of this change is to ensure that the two B train NS heat exchangers that may possibly be in service for this pump test are isolated prior 168 9 9
~
,y , ,
to. shutting down the pump being tested. This will prevent.any chance of puop runout occurring for the
' remaining in-service pump is that pump is alone supplying the~ station load. ,
This change in no way increases the probability of an accident occurring, whether or not it has been !
evaluated in the FSAR. RN will always be supplied to all components, so'nc htaracthn with the reactor or supporting systems 's ptesent.' The t consequences of an accident prev;: sly evaluated in the FSAR are not increased, since the opposite train-0 of RN will be in service and supplying cooling water-
'to required' components.
This change does not increase the probability of nor -
consequences of a malfunction of safety related equipment. RN cooling water will be supplied.at all .
s times to all equipment on both essential headers and. >
nonessential headers. This change decreases the_ -
probability of a malfunction of equipment important-to safety by reducing the chance of pump runout. No new equipment failure malfunctions are created as a result of this change.
The train of RN being tested will be inoperable at the time of the test. Technical Specification 3/4.7.4 allows'two train related R_N pumps to be out i of service for up to'72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in ,
Modes 1, 2, 3, and 4. This change will not compro- ,
mise the operability of the opposite train pumps.
Therefore, the margin of safety as defined in the bases of Technical Specifications is not reduced.
The second.part of this change is to allow Perfor-mance technicians to remove and install the access ;
plates to the-upper-bearing vibration monitoring points. In the past, a Standing Work Request was used to allow Mechanical Maintenance to perform this "
activity. These SWR's have been deleted so Perfor-mance will now remove these plates. This has no c' effect on the test method, test results, or pump operability. :
The third part of this change merely corrects.a typographical error in the nomenclature of a valve.
The valve number as specified in the procedure is correct. This also has no effect on the test method, test results, or pump operability.
PT/2/A/4600/11, Neutron Noise Data Acquisition: .This change adds
. Change #5 provisions for the monitoring of two additional o Incore Therocouples (subject procedure presently allows for the monitoring of 13 of the.Incore T/C's) 169 e
a .
F w .
r for the purpose of assessing the behavior of ob-m served reactor coolant flow anomaly. None of these-
+
T/C's are Safety Grade, so Tech Spec 3.3.3.6 is not i affected by this change. . Indication of core exit ;
a temperatures by these thermocouples will not be affected by visicorder trending via test-leads. :
installed in the ENA Cabinet in the Control Room. !
0AC Program Nuclear.11, Thermocouple Map, will still
~
receive valid process' indications for the purpose of calculating core quadrant power tilt and FAH (as.
inferred by T/C indications). .The basis for the calculations of the Thermocouple Map. program is- :
stated-in FSAR Section 4.4.6.1 which refers to the.
Incore T/C's as a backup for the Incore Flux Mapping System.
,~
Neither the probability nor the consequences of any accident or safety significant equipment malfunc- .t tion, either analyzed or unanalyzed by the FSAR, will be created by this change. The Incore T/C's themselves (described in FSAR Section 7.7.1.9.1) are not affected in any.way by this change. The margin'
- of- safety of Tech Specs: will .not be reduced since-this change has no impact on any Tech-Specs.
i PT/0/A/4400/08, RN Flow Balance for Degraded Mode: The purpose of 3 Change #39 ' modification CN-50386 is to add 2 simplex strainers :
in parallel with the normal RN Pump Lube Injection-Strainer. .This modification will allow the strainer to be removed from service for cleaning without making that train of RN inoperable. This retest will verify that' the modification had no affect on .
normal strainer' flows, and will also' verify that.
minimum acceptable flows can'be obtained using the
-new strainers. 3
\
u P This charige in no way increases the probability of
an accident occurring, whether or not it has been .i evaluated in the FSAR. Cooling water to all inservice components.is supplied at all times, so no. :
interaction with the reactor or supporting systems ;
'is present. The consequences.of an accident previ- j
.ously evaluated .in the FSAR are not increased, since !
the opposite-train of RN will be.available to supply cooling water to required components it necessary. -
This change 'does not increase the probability 'of nor consequences of a malfunction of safety related equipment. Before valving out any strainer, the strainer being tested will be placed in service.
This will ensure that adequate bearing injection and 4 i4 motor cooling water flows are maintained at all a times. .No new equipment failure malfunctions are created as a result of this change.
170 l
'The. train of RN being tested will be inoperable'at the-time of the test. Technical Specification 3/4.7.4 allows two RN pumps to be out of service for-
.up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with both units in Modes-1, 2, 3, and-
- 4. This change vil.1 not compromise the operability.
of the opposite train pumps. Therefore, the margin of safety as defined in'the bases of Technical Specifications is not reduced.
OP/0/A/6400/06F Nuclear Service Water System F)ush Procedure: This procedure is designed to flush sediment from the KC, KD, and NS Hxs and maintain operability of the RN-System. The heat exchangers are flushed to maintain -
them as clean as possible since RN is supplied by a raw water source (Lake Wylie or SNSWP). 'The proce-dure ensures runout will not occur during flush by allowing 'no other component on the RN System to be flushed at the same time (Flush one component at a time)'and requiring sufficient RN pumps to be operable during'the flush. Therefore, no change-is required to the FSAR by-this-procedure.'
PT/2/A/4550/03B Spent Fuei Pool Inventory: This inventory test is not required by Tech Specs or mentioned in-the FSAR, but it is required by 10CFR70.51. CNS is committed to perform this test every six months per APM 3.9.5.1 and Station Directive 3.9.1. This change simply reduces the mandatory requirement in the procedure to do a serial number identification of -
inserts and makes it optional. The procedure will still perform the physical inventory of SNM fuel-assemblies.
This change does not affect any equipment important to safety and does not increase the probability or consequences of an accident. The margin-of safety as mentioned in Tech Specs will'not be reduced.-
PT/1/A/4200/09A, Auxiliary Safeguards Test Cabinet Periodic Test:
Change #129 Neither the probability nor the. consequences of an.
accident previously evaluated in the FSAR will be increased, nor will the possibility of an accident different than any already evalvated in the FSAR be created by this change. The KC non-essential header isolation valves on the train W ing tested close, and the KC supply to the ND heat exchanger-opens.
All of the valves are in their " safety" position. -A step is provided in the procedure to ensure that the opposite train of KC is operating to provide cooling water to the non-essential header. When 1KF103A (1KF101B) is tested, the opposite train valve is ensured closed to prevent draining the FWST. Valve 1NI185A (1NI184B) is deenergized to prevent the valve from cycling ano Waininc . ..
1-tainment. Neither the p 7habj e rh- ,
i l
171 l
[. T consequences of a malfunction of equipment important to safety will be increased by this change. KC flow will be maintained to components during this test.
p The margin of safety as defined in the bases to Tech-Specs will not be reduced by this change.
OP/0/A/6400/06C Nuclear Service Water System: See 50.59 evaluation.
for NSM CN-50386.
OP/1/A/6450/02 & Annulus Ventilation System: PIR 0-C-89-278 concern-OP/2/A/6450/02 ing the inability of .the Catawba Annulus Ventilation Systems to provide adequate negative pressure in all Annulus regions at all times, resulted in lowering the pressure setpoint to ensure that -0.5 inches of water vacuum is maintained. Exempt Change CE 2466 (Unit 1) and CE 2467 (Unit 2) reduced the VE dif-ferential pressure setpoint on the controller to
-1.5 inches of water from -1.0 inches for both VE trains of both units.
Additional failure considerations for VE resulted in j a maximum negative pressure setpoint of -2.0 inches of water. This prevents excessive annulus flow reaching the Control Room via a recirc flow path and allowing an excessive skin dose to be received by Control Room Operators. ,
i These changes have been reviewed and approved by Duke Power Design Engineering Department and, there-fore, do not constitute an unreviewed safety ques-tion. There are no effects on existing equipment except for slight position changes for the exhaust damper while controlling at the new setpoint.
I The FSAR descriptions of the operation of the VE sydet1 is undergoing revision by Design Engineering
, to more accurately reflect the system setpoint change.
PT/1/A/4450/03A, Annulus Ventilation System Operability Test: PIR
.PT/1/A/4450/038, 0-C-89-278 concerning the inability of the Catawba PT/2/A/4450/03A & Annulus Ventilation Systems to provide adequate PT/2/A/4450/038 negative pressure in all Annulus regions at all times, resulted in lowering the pressure setpoint to ensure that -0.5 inches of water vacuum is main-tained. Exempt Change CE 2466 (Unit 1) and CE 2467 (Unit 2) reduced the VE differential pressure setpoint on the controller to -1.5 inches of water from -1.0 inches for both VE trains of both Units.
Additional re ure considerations for VE resulted in a maximum negs ive pressure setpoint of -2.0 inches of water. This prevents excessive annulus flow reaching the Control Room via a recirc flow path and
i J
w
~'
allowing an excessive skin dose to be receivdd by '
' Control Room Operators. ;
These changes have teen reviewed and approved by ,
Duke Power Des' "agineering Department and, there-fore, do not coni Itute an unreviewed safety ques-tion. There are no effects on existing equipment except for slight position changes for the_exheust ,
damper while controlling at the new setpoint.
The FSAR descriptions of the operation of--the VE 1 system is' undergoing revision by Design Engineering to more accurately reflect the systea -setpoint
. change. -l PT/1/A/il50/12B Isothermal Temperature Coefficent of Reactivity-Measurement (EOL): :The only substantive change in the reissue of this procedure is the_ replacement of the Westinghouse Nuclear Design Report (NDR) with the Duke Power Design Engineering Startup and 1 Operational Report (SOR) as the' core design data resource for completing the calculations involved in the test. The Duke Power SOR is as rigorously ,
reviewed and qualified a document as the Westing-house NDR, making this replacement permissible.
Calculation of the Moderator Temperature Coefficient
. at End of Life (< 300 ppmB critical boron concen-tration) for the purpose of ensuring compliance with Tech Spec 3.1.1.3 is performed in exactly the same ,
way, .the data to_ do so is simply obtained from a different reference source. Due to thorough safety analysis of the Duke Power SOR methodology and ,
results, neither the probability nor the conse-quences of.any accident of safety related eauipment malfunction (either analyzed or unanalyzed) will be-increased by this change. The margin'of safety as defined in Tech Spec bases will be maintained ,
through verification of compliance with Tech Spec 2.1.1.3 via acceptable methodology.
TT/0/A/9100/49 Control Room Area Ventilation Auto-Start During ECCS Actuation: Design study CNDS-189/00 is reviewing 'I Catawba's ECCS actuation system which requires both j trains of the VC system to start on a Safety Injec-tion (SS) signal and evaluating'an appropriate resolution to reduce the post-DBA sound level in the Control Room (CR). The purpose of this temporary a procedure is to obtain data on the VC System before and after an SS signal has been received to aide in the resolution of'the referenced Design study.
Various data will be taken on the VC system before ,
and after the SS signal has been initiated. In 173 .
, q E ,
.(
order to initiate an SS signal on E, train of VC, a=
jumper will momentarily be placed in the B train-
, control cabinet.- The momentary placement of this- '
jumper will allow resetting 0; train of.the VC system
? at any time during the~ test without having to reset-the D/G sequencer first. Since the B train chiller will not start on the SS signal, the procedure requires that CR temperature be monitored continu ;
-ously after the SS signal has been initiated until B train is reset. If the temperature in the CR approaches 85 deg F, after initiation ofnthe SS -
signal, the-procedure' requires that B train of VC be' f reset immediately to reduce CR temperature. Tech Spec requires the CR temperature to be 5 90'de'g F so a conservative value of 85 deg F was chosen. If any .'
unusual conditions arise with Unit 1 or Unit 2 during.the post SS signal operation, the test may be aborted and A traisi of VC returned to the normal
. operating alignment. ,
Train A of VC is required to be operable by ths procedure and will be the selected train of opera- ,
tion before the SS signal. Initiation of the SS. '
signal will align the VC system to the post-DBA operating alignment so that the margin of safety as defined in-the bases of Tech Spec will not be ,
reduced. All data will be taken while the system is 1 operating and the measurement of this data will not affect the operation OR operability of the VC system. Placement / removal of all jumpers and .
installation / removal of all applicable test equip- l ment will be independently verified within_ Sections -
12.0 and 13.0 of the procedure. -For'these reasons' and the ones stated abovu, this procedure does not increase the probability or. consequences of an accident previously evaluated in the FSAR OR in- 4
-crease the probability or consequences of a mal-function of equipment important to safety. Also, this procedure will.not create the possibility of an accident or malfunction of equipment important to .
safety different than'aiready evaluated in the FSAR.
Therefore, this procedure does not create an unreviewed safety question.
MP/0/A/7600/23A BIF Butterfly Valve Removal, Replacement, Corrective 4 Maintenance (Handwheel' Operated): This procedure provides a method for: disassembly, inspection, '
reassembly, and corrective maintenance for BIF-Butterfly Valves (HW ONLY). Technical information was obtained from manual CNM 1205.02-111 and draw-ings CNM 1205.02-32, 33, 38, 39, 41, 42, 49, 51, 53, -
60, 64, 174, 223. This procedure also contains information specific to Catawba. '
I
-174
(-. ,
f The purpose of this evaluation is to desc;ibe the C changes made to MP/0/A/7600/23A as part of the proceduro upgrade process and'the identified tech-nical enhancements. While many of these changes are not'significant in content and_are editorial in nature, some 'chnicci information was added.
The following is a summary of the changes made to i this procedure:
- Provided more detail on the seat preparation y ano installation. i
- Addition of scribe marks t' nositively indicate I the disc position inside t e aive.
, s
- Addition of specific instruct'.ons on the method of replacing the disc on valses with a retain- -
ing ring;
- Addition of steps to pressure test the valve after repairs and/or before return to. service.
- Provided more detail on the setting of the operator mechanical stops.
- Addition of steps to install an upgra kt position indicator and to verify that it is accurately set. -
- Addition of steps to record specific valve information to maintain a complete history for_
the valves.
- Addition of enclosure denoting torque values retrieved from the 0/L drawings. ,
The purpose of this procedure is to correct and _
improve the performance of these valves within their original design requirements and specifications.
l- The FSAR and Technical Specifications were reviewed, .
L and neither will be affected by the changes de-scribed above. The probability of an accident or malfunction previru ly~ addressed will not be in-creased. No unrevit:wed safety questions are in-volved.
Mp/0/A/7600/23B BIF Butte. fly Valve. Removal, Replacement Corrective Maintenance.(Electric Motor Operated): This price-dure provides a method for disassembly, inspec ion, reassembly, and corrective maintenance for BIF Ntterfly Valves (EMO ONLY). Technical information '
was obtained from manual CNM 1205.02-111 and 17_5 i
t
m- ,
Y y .
drawings CNM 1205.02-33, 35, 39, 42,-56, 58, 62, 66, 68, 148, 149,-176, 223, 225, 243, 296. This proce dure also contains information specific to Catawba..
The purpose of this evaluation is to describe the changes made to MP/0/A/7600/23D'as part of the procedure upgrade process-and the identifted tech-- i nical enhancements. While many of these changes are. ,
not significant in content and are editorial in nature, some technical information was added.
The following is a summary of the changes made to this procedure.
P
- Provided more detail on the. seat preparation and installation.
- Addition of scribe marks to positively indicate the disc position inside the valve.
- Addition of specific instructions on the method j of replacing the disc on valves with a retain ' ,
ing ring. F
- Addition of steps to pressure test the valve after repairs and/or before return to' service.
- Provided more detail on the setting of the operator mechanical stops. ,
- Addition of steps to install an upgraded - .
position indkator and to verify that it is accurately set.
- Addition of steps to record specific valve information to maintain a complete history for q the valves, i
- Addition of enclosure denoting torque values ret b .ed frcm the 0/L drawings.
The purpose of this procedure is to correct and improve the performance of tht.se valves within their -!
original design requirements and specifications.
The FSAR and Technical Specifications were reviewed, and neither will be affected by the changes de-scribed above. The probability of an accident or malfunction previously addressed will not be in-creased. No unreviewed safety questions are in-volved. i PT/2/A/4200/13K CA Valve Inservice Test (CS): The purpose of this procedu m is to perform IWV valve stroke tests of
, i t'
176
c
- 2CA149, 2CA150, 2CA151, and 2CA152. These valves are normally tested during Cold Shutdown.(Modes 5, 6, or No Mode). However, upon review of past procedure' changes, it_was found that temporary changes were written when these valves had mainte-nance performed on them and were retested in other modes, usually at the cnd of an outage while. Unit 2
- was attempting startup (Modes 3 and 4). Testing these CA valves involves stroking each valve froin
- open to closed, thus'. isolating flow to the affected steam generator upper nozzle." However, as long as
{
the lower nozzle is receivin 85% or.less of total
. rated CF flow to the steam r3enerater (total CF flow to the affected steam generator at 100% rated
- power), the CA valve may be isolated.' Although ultimately, these CA valves.could be tested in all J modes, this procedure limits testing to the follow-
[ ing: Modes 2, 3, 4, 5, 6, and No Mooe. Addition- ,
B .11y, when isolation occurs during testing, if steam
- s being produced in the steam generators, and
_ y' Auxiliary Feedwcter (CA) is the sole supplier of feedwater, t% water level of the isolated steam generator could become low enough to exceed the Low Low level trip point, causing a CA autostart signal.
Precautionary steps and measures are being. inserted E in this retype to warn against low steam generator E levels which could result, should these CA valves be t left closed for a prolonged period.
F Since the flowpath of 2CA149, 2CA150, 2CA151, and 2CA152 may be isolated during Modes 2,,3, 4, 5, 6,-
or No Mode, no CA valve is being made inoperoble as a result uf this: test. As a result, neither the probability of nor consequences of:any' accident
- previously evaluated in the FSAR'are increased, No accident different than those already evaluated in the FSAR are created. The probabi.lity of and E consequences of a malfunction of equipment important R to safety different.from those already evaluated in K ,
the FSAR is not created. The margin of safety as defined in.the Technical Specifications Bases is not reduced.
MP/0/A/7150/60 Pall-Trinity Filter Removal and Replacement: This safety evaluation is for the reissuing of g~ MP/0/A/7150/60 changes 0 to 12 incorporated, pre-pared on 11-01-1988. This procedure has been upgraded. The changes to the procedure are as follows:
r Section 1.0 Revise for clarity.
Section 2.6 Added sections of Power Chemistry Guide.
177 mr
r 7 7 ,
k o
4 p
Section 4.1.1 Upgraded to present format.
Section 4.2.1 Upgraded to present format.
Section 4.3 Added-statement. :
E 3ection 6.0- -Added steps.6.1 and 6.2 and sign-off to;6.3 Section 8.0 Added 8.1:4 through 8.3.1 Section 9.0 Upgraded _to present format.
F 'Section 11.0 All sign-offs were upgraded to
-! -present-format and step 11.3.6 was added.
Section 12.0 Added statement. ;
Enclosure 13.1 Revised to reflect changes ;
made above.
A full USQ evaluation is required because this procedure was changed significantly and .it is '
described in the FSAR. FSAR Section 13.5.2.2.1
,ilaintenance Procedures) stetes that maintenance procedures are: required for safety related equip . ._'
ment. Since this rewrite, as described above, changes this procedure in a significant manner, a full USQ evaluation must be performed.
Tech Specs 3.11.'1.3 and 3.11.1.1 may be affe'cted by this procedure. -Operations has the res' isibility and the procedures for compliance with .ese Tech-Specs. Maintenance will be performed on these filters'when Tech _ Specs allow, per Operations'- ,
procedures. This rewrite will c19rify'and assure that maintenance activities will return the filters to as-designed conditions.
i The changes made by this rewrite have been reviewed a anainst approved vendor _ manuals, design-documents, ,
-and station procedures to ensure that the corrective '
maintenance controlled by'this procedure will return the filter to as-built /as-designed condition._ These actions will ensure the filter's compliance with t FSAR accident. analysis. Since the filter will be returned to as-designed condition's', the possibility, l
consequences, or probability of a' malfunction will be reduced. Therefore, no USQ exists.
L MP/0/A/7600/37 Borg Warner Pressure Seal Check Valve corrective 4 Maintenance: This procedure perform:, cirrective 4m
~
] , ,
4 4
maintenance on Borg Warner pressure seal check valves. Instruction nianual CNM-1205.00-0323 and CNM-1205.00-105S provided technical .nformation for the development of the procedure. Plant-specific-information that applies to Catawba Nuclear. Station.
was also included in its A velopment.'.
This evaluation is for changes made during the-procedure upgrade process. .These changes are in the procedure format and are not sig'nificant-in content.
The following are_ technical changes that.were made ,
to this procedure during this rewrite, k
- Step note, were added to ineure that the bonnet to body joint ~ is prcperly y assembled.
Added note to Step 11.4.14 to stress the importance of proper bonnet te arevent
~
j seat leakage. I
{
Added note to ltep 11.4.22 to insure l bonnet is snug against seal ring before j starting torque passes.
}
- Steps were added to record and inspect for I check valve wear related to SOER 86-03.
Added Step 11.3.3 for Hinge Pin wear.-
f Added Step 11.3.4 for Disc Stud wear.
g.
Added Step 11.3.5 for Disc Stud to' Disc j Stop inspection. .q
- Added Step 11.4.23 to retorque body to bonnet joint at operating temperature and pressure.
]
This evaluation is for changes made during.the procedure upgrade process. These char.ges are in the pro edure fermat and are not significant .in content. 1
, The Catawba FSAR and Technical Specifications have l been reviewed tr<d are not affected by this-proce-dure. This procedure applies to valves used in l various plant applications. This procedure will be used to correct and improve the performancc. of-the valve and will maintain the valve within its origi- .,
nal design requirements and specifications. The j safety standards previously specified within the FSAR and Technical specifications will be main- ,
tained. The probability of an. accident or malfunc-tion previously addressed in the FSAR wi . not be 1 increased. No unreviewed safety questions are involved. -l 179 L d
s..
PT/1/A/4200/01G Mechanical Penetration Bellows Integrity Test: The >
purpose of procedure PT/1/A/4200/01G is' to verify the integrity of the dual ply bellows assemblies on containment penetrations.
System Description
A mechanical penetration bellows assembly provides ,
'the means for routing a process pipe through an :
aperture in the' containment pressure barrier while ;
maintaining the integrity of the pressure barrier.
In addition, the bellows assembly allows for thermal expansion / contraction of the process pipe due-to ,
heat loads from the process fluid or from the- ,
containment environment.
The SM and CF mechanical penetration assemblies each contain four dual piv bellows, while the' fuel o transfer tube contairis two. The remaining mechani-cal penetrations each contain one dual ply bellows.
Tubing is connected to each outer bellows to allow leak rate testing the volume between the ir.ner and-outer bellows. The inner (safety-related) bellows provides tne pressure boundary for containment t integrity. The outer bellows is not safety-related; - -
it merely provides a pressure boundary for leak rate.
testing the inner bellows. Leak rate testing between the bellows is performed using the makeup flow method at 4.0 psig. The reduced test pressure is necessary to prevent damaging the inner bellows, which could not withstand full test pressure from tne outside. If any leakage is detected at the reduced pressure, a leak rate test is performed with -
pressure on'the containment side of the inner bellows at.15.4 psig, and the measured leakage rate is added to the total Type B and C leakage rate.
Safety Analysis This procedure is performed in Mode 5, 6, or No Mode !
(containment integrity not required), so the proba-bility/ consequences of an/ accident previously evaluated in the FSAR will not be increased, and the possibility of an accident which is different from any already evaluated in the FSAR will not be created. L i
During the reduced pressure test, the only change to the system is the re m al of the test connection cap :
on-the tubing to the outer bellows. However, since i the tubing is connected to the outer bellows, the test connection cap is not required for containment integrity. Overpressurization during the reduced 180 m ,
= .
i ,
pressure test.could result in damage to the ' inner 1 bellows; However,__ damage to this containment pressure barrier wou'd be immediately detected via-an increase in'leakat. as. indicated by the Volumetrics Leak Rata Monitor.
If leakage is detected during the reduced pressure test, it is n*ct possible to determine if the leakage is from the ..< < r or outer bellows. Therefore, the mechan 1 cal penetration bellows assembly is declared inoperable, and the inner bellows must be tested from inside containment at full accident pressure.
(Zero leakage during the full preesure test would
. indicate that the leakage measurea during the reduced pressure test was from the outer bellows.)
s During the full pressure' test, the outer bellows is vented to atmosphere (via test connection).to assure full differential pressure across the inner bellows.
In addition, when full-pressure testing a penetra- $
tior, assembly containing more than one bellows, all outer bellows in that penetration are vented. .This will preclude a pressure build-up in the volume between the dual ply bellows, which could result in_ ,
damage to the inner bellows _when depressurizing the test-volume. ,
'In order to test the bellows assemblies from inside containment, a plug is installed inside containment to provide a pressure boundary. Since the plug may inhibit bellows assembly movement due to therrnal. y expansion of the process pipe,-the plug is removed. ,
prior to entering Mode 4. For these reasons, the probability / consequences of a malfunction of equip-ment important to safety will not be increased, and ;
no margin of safety will be reduced.
PT/0/A/4450/08C, Control Room Area Ventilation Flow Balance: The -
Change #2 purpose of this restricted procedure change is to ensure operability of Train A VC while clos-ing/ opening 2PFT-MVD-2. It has been found that when j B Train of VC is in operation, air was leaking past 2CR-D-9 and ICR-D-10 into the outside air line and entering'2CRA-PFT-1. This leakage through the
' p" opposite train affects the ability of the operating-train to pressurize the CR. In. order to wori. on ICR-D-9 and ICR-0-10, the CR recirculation line was isolated to ensure B Train Operability while per-forming the work. Refer to change 1 to this proce-dure. Change 2 has been written to repeat the same process on A Train since Inspection o' 2CR-D-4, l
2CR-D-9 and 2CR-D-10 indicates that each damper is leaking and not sealing properly.
181 x
7 e
g -(
'In an accident condition, one primary purpose of the VC System is to provide uncontaminated. filtered air to the Control Room (CR) and to pressurize the CR to ensure no inleakage of contaminated air. Outside
+
air, taken normally fror two intakes, is used to pressurize the CR and the'CR area.. Tech Spec requires less then 400b cfm of-total outside air. ,
During an accident, one (or both) of these intakes '
~
may isolate due to high radiation so VC must be able to pressurize the CR with either intake isolated (if both intakes isolate, the Operator opens the-least ;
contaminated one). This outside air and recircula--
o tion air from the CR pass through a filter unit a consisting of HEPA filters and a carbon absorber bed. The flow through this filter unit should be close to the design flow to ensure proper filtration of the, air. The above mentioned flows and CR pressurization are some of the critical parameters
- as m ed in the Safety Analysis for VC.
This change requires continuous monitoring of the critical parameters to ensure that they stay within ;
their safety limits while closing / opening 2PFT-MVD-2. System MVD's may be adjusted to maintain these parameters. 2PFT-MVD-2-will be closed / opened in. increments and critical parameters will be verified each time the MVD is moved. Final position of P,vT-MVD-2 will be the same' as the "As Found"~
pos'uen to ensure that the B Train balance will not be iffected. Therefore, the margin of safety as defined in the bases of Tech Spec will not be reduced. Installation / removal of test instrumenta-tion, securing the cap at the discharge of ICRA-PFT-1 and securing all adjusted MVD's are independently verified within Section 12.0 or 13.0 of this change. Also, the final position of all adjusted MVD's will be markea and the method of marking will be documented on Enclosure 13.11. For these reasons and the ones stated above, this procedure does not create or increase'the probabil- "
ity of a malfunction of equipment important to safety OR create or increase the consequences of en i accident. Therefore, an unreviewed safety question-does not exist.
-PT/1/A/4450/05B, Containment Air Return Fan 28 and Hydrogen Skimmer Change #12 Fan 2B Performance Test: A change was made to this <
procedure to allow using Digisnap Model DSA-1000 clamp-on ammeter for the measurement of motor currents. The acceptance criterion for motor ;
currents were adjusted to encompass the additional '
error associated with the new ammeter (see Error ,
analysis on pages 3 and 4 of this change). A change was made to allow a VQ release during the 182 ,
p e s
W ',
p p
y
( f f~ performance of Secti' i* During:the performance
, y~ of Section 12.4, the ,,etwn valve (2VX2B) is closed -
y'
~
when the HSF-2B faa a running and no. air is moved.
Therefore, airborne contamination in containment a will.not be increased by Section 12.4. Other i changes (not including the change. discussed below) were made to improve' clarity, delete confusion within the. procedure and correct typographical- l errors. ,
A change was made to open then close the breaker for- ,
the'HSF-2B if.the fan is operated by the procedure.'
Problems were encountered with the breaker on HSF-2A-tripping and a recommendation was made to cycle the breaker associated with HSF-2B after running the fan a to ensure that the breaker trip latch lever is fully.
reset (see memo to file on pages 6 and 7 of this change). After the breaker is re-closed, the indication on 2MC4 is checked to verify that power has been returned to HSF-28. .The procedure requires
-that the HSF-28 is declared inoperable during this '
test and the HSF-2A will remain operable during this test as required by Tech Specs. Cycling of the breaker will decrease the possibility of a malfunc-tion of equipment important to safety. Open- .
4 ing/ closing the breaker -and verifying that power is
. returned af ter re-closing the breaker is indepen- ,
dently verified within Section 12.5 of the proco- 4 dure. Therefore, the margin.of safety as defined in ,
the bases of Tech Specs will not be reduced. !
- [
The order of steps and alignment of systems was not changed. For.this reason and the ones stated above, this procedure change does not increase the conse-quences or probability of an accident OR create or increase the probability of a malfunction of equip-ment important to safety. The-possibility of an accident different than already evaluated will not be created by this change. Therefore, an unreviewed a
safety question does not exist. :
PT/1/A/4200/17, BB Valve-Inservice Test (QU): This change involves Change #10 the method of recording stroke time for valve 4' 1BB-56A. This valve has a Technical Specification required stroke time of.10 secc,nds per TS 3.6.3. ;
That spec also says timing will be pursuant to TS 4.0.5 which references ASME Section XI code. The code (IWV-3413(b)) allows stroke timing to the nearest second.
l
, 'After discussion with station personnel, General Office Compliance and Design personnel, NRC resi- !
dents and NRC Region Il personnel, timing to the 183
o RJ nearest second would have no effect on'the proba-bility of-an accident or malfunction of equipment, a
? 1BB-56A has shown no degradation (steady increase in 9 stroke time) over the past two years of_ testing per J IWV. Should degradation occur and the stroke time-increase to > 10.5 seconds. The valve would be declared expectable and corrective action initiated.
No new possibilities.of accident or equipmen't malfunction are created.
~The margin of safety is not reduced.
.MP/0/A/7600/37 Borg Warner Pressure Seal Swing Check Valve: This change to MP/0/A/7600/37 is to incorporate a caut%n concerning the removal and reinstallation of the bonnet-retainer. This caution is the result of NRC Information Notice 89-62. As described in the information notice, gross back leakage eu occur due to improper location of the bonnet retainer. -The maintenance procedure now requires the mechanic to contact Engineering Services prior to disassembling the valve. This is a precautionary measure to ensure the same problems described in-the informa-tion notice are avoided. In effect, this change will act to reduce the probability of equipment malfunction. Consequently, no.unreviewed safety questions are create ' and the margin of safety as defined in the bases to the Technical Specifications will not be reduced.
IP/0/A/3817/12 Calibration Procedure for Barton 763, 764, and 386A Pressure Transmitters: See 50.59 evaluation for exempt change VN CE-2223.
PT/0/A/4450/08C Control Room Area Ventilation Flow Balance: The.
purpose of this procedure is to optimize both Trains of the VC system flows to the Control Room and the Control Room Area. In an accident condition, one primary purpo:e of the VC System is to provide uncontaminated filtered air to the Control Room (CR) and to pressurize the.CR to ensure no inleakage of-contaminated air. Outside air, taken normally from two intakes, is used.to pressurize the CR and the CR area. During an accident,:one (or both) of these intakes may isolate due to high radiation so VC must be able to pressurize' the CR with either intake isolated (if both intakes isolate the Operator opens theleastcontaminatedone). This outside air and revirculation air from the CR pass through a filter unit consisting of HEPA filters and a carbon ab-sorber bed. The flow through this filter unit should be close to the design flow to ensure proper
~
filtration of the air. Flow existing this filter 184
1 unit is split as necessary_to pressurize the CR and x the CR area. Updated criteria-for the split of these two flows has been developed by Design-Engi-neering to ensure proper pressurization of the CR and CR areas for all ope ~ rating alignments. Sections of this proceduro allow taking measurements and adjusting Manual Volume Dampers (MVD's) as necessary to optimize all of the above mentioned flows and CR pressurization.- If a MVD is moved in one section that could affect the flows in another section, th9 procedure requires that the affected section is-performed OR the sections are-performed concurrently and' final data is not taken-until all MVD's have been adjusted.
Other sections of this procedure allow taking measurements and adjusting MVD's to ensure other areas of the VC system are operating as-designed.
Total supply flows to the CR, CR area, switchgear rooms and exhaust flows from the battery room will be verified to ensure proper operation. Acceptance criterion for these sections was developed-using CNTC-1578-VC series of Test Acceptance Criteria Sheets from Design Engineering.
All measurement of flows and adjustment of MVD's are performed while the VC System is in its normal operating alignment. When performing _the-sections that affect or may affect safety related flows due
, to MVD adjustments, the Train of VC under. test will.
be declared. inoperable. The opposite Train of VC will remain operable during this period as required by Tech Specs.. Therefore, the margin of safety as defined in the bases of Tech Spec will not be reduced. Installation / removal of_ test'instrumenta-tion, securing duct access' doors and securing'any adjusted MVD's are independently verified within o Section 12.0 or 13.0 of the procedure. Also, all adjusted MVD's'will be marked and the method of marking will be documented on Enclosure 13.11. For these reasons and the ones stated above, this procedure does not create or increase the probabil-ity of'a malfunction of equipment important to safety OR create or increase the consequences of an accident. Therefore an unreviewed safety question does not exist.
MP/1/A/7150/42 Reactor Vessel Head Removal and Replacement: This procedure performs removal / replacement activities required on the Reactor Vessel Head during refueling outages.
This evaluation is for changes made during the procedure review following 1-E0C-3 outage. These 185 w - _____ _ _ _____ _ _- _ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ __ _ __ -
O, i F> j W.
?
changes were required to clarify steps for better
- 1 understanding a-d add manufactures information on
.,' new equipment. See procedure changes listed below. '
L
- 2.1.6 - Added CNM-1144.28.0037-001, Instruction manual for R.V. Nozzle Inspection Hatch' Covers. -(
.l
- 8.2 - Added 0-300 torque wrench.
- 8.3.12 and 8.3.13 . Reduced tygon tubing and rope length.
L
- 10.2 - Added note on head shielding blanket installation. Moved from Step 10.6.1.
- 11.3.2 - Added 2744 SWR to step.
- 11.3.8 thru 11.3.8.2 - Deleted these steps due ,
to new mechanical seals.
- 11.4.24.1 - Increased torque on stud hole plugs a
, to 60 ft. lbs.. This recommendation from plug manufacture to reduce leakage.
- 11.5.33 and 11.4.37.1 - Revised to add-instal-
'lation instructions for mechanical sealing nozzle covers per manufacture's manual.
- '11.5.14 and 11.5.15 - Changed wording to require-inspection on head lifting rig.
- 11.5.23 - Added caution step for stud hole. t cleaning tool brush rotation. '
- 11.5.33 - Changed wording for operations sign-off to allow tensioning:to begin and :
provide a time for expected Mode 5 requirements. a, j '* 11.5.69 thru 11.5.72 - Revised to add removal t instructions for nozzle covers. ;
l 3
- Enclosure 13.6 - Added nozzle port cover
- l. locking sequence. "
- Enclosura 13.1,-Paga 14 of'23 - Relocated Step-
- i. 11.5.39 for clarity.
This procedure will be used to maintain the Reactor Vessel Head in it's original design requirements and specifications. The Catawba-FSAR and Technical' Specifications have been reviewed and are-not !
affected by this procedure change. The probability '
of an accident or a malfunction previously addressed will not be increased nor will any new'unreviewed
~
safety question be involved.
18L6 N_
. m l
l PT/1(2)/A/4550/03B Spent Fuel Pool Inventory: This~ inventory procedure is not required by Tech Specs and'is not mentioned ;
'in the FSAR, but it is required by 10CFR70.51. CNS. -
is committed to perform this procedure every six months per APM 3.9.5.1 and Station Directive 3.9.1. .
This procedure while satisfying the Special Nuclear' .
Material-(fuel) inventory in the Spent Fuel. Pool F
does this without placing any equipment in an'-
unusual position from which the are designed. No ,
Tech Specs are violated by the use of this:proce-dure.
PT/1/A/4200/41B, Containment Air Release and Addition Isolation Valve u Change #19 Leak Rate Test: The VQ System provides a means of controlling the containment pressure between'-0.1 psig and +0.3 psig during normal plant operations. "
To decrease containment pressure (air release mode), ;
tne VQ fans take suction from containment through <>
IVQ2A and IVQ3B and discharge to the unit vent <
through 1VQ10. To increase containment pressure (air addition mode), vacuum IVQ158, IVQ16A, and IVQ13 are opened to allow the vacuum inside con- i; tainment to suck air from the Auxiliary Building. .
This procedure change will allow leak rate testing of valve IVQ16A from outside containment, utilizing 10978 SWR, The SWR will install a plate.'in place of
- a. flow-restricting orifice to provide a pressure-boundary for the leak rate test. The plate wil1~
block the_ recirculation flow path-from the discharge of'the fans through IVQ15B and IVQ16A. -However, 3 this flow path is not utilized in either the air release mode on air addition mode. (Reference VQ' Operating Procedure OP/1/A/6540/17). Therefore, the.
VQ System will still be functional while this blank orifice plate is installed. 1 o
Furthermore, the only safety function of the VQ System is containment isolation (1VQ15B and IVQ16A - ,
both receive an St signal), which is not'affected by this test. Therefore, the probability / consequences of any new or previously analyzed FSAR accidents or ,
safety-related equipment malfunctions are not l increased, and the margin of safety is not reduced.
l
-0P/2/B/6300/01, Turbine Generator: Retype #4 of OP/2/B/6300/01 Retype #4 (Turbine Generator) includes the following changes:
- 1) Changed Analog Computer Points in Step 2.5 of Limits and Precautions.
a .,
- 2) New Step 2.22 was added to Limits and Precau- i u
tions to inform the' operators as to the 9
+
182
s, 4
[U I
[
expected response when~the turbine is reset'and the. generator are in the indicated positio.n.
- 3) Added location for the generator and exciter' field ground detector relays (Step 1.20 of Enclosure 4.1).
4). New Step 1.29 was added.to Enclosure 4.1 to ensure that the " Throttle Pressure Limiter" is "off".
- 5) NOTE was added prior to' Step 2.10.7.4 of; Enclosure 4.1 to inform operators that the control valve may not initially,come full open in.shell warming mode due to the throttle pressure limiter signal.
- 6) New Step 2.10.7.8 was added to Enclosure 4.1 to 1
-verify valves 2SM-21, 25, 29 and 33 close after, 1 the shell warming pushbutton is depressed.
- 7) NOTE was added after Step 2.10.7.8 and Step !
2.16.3 of Enclosure 4.1 to inform operators as to how valves 2SM-155, 158, 161 and 164 may be. !
positioned in shell warming mode.
I J
- 8) NOTE prior to Step 2.10.8 of Enclosure 4.1 was l changed to not require computer point D2802 .
(Turbine Status) to indicate "NOT' TRIPPED".
q l 9) New Step 2.10.9.1 was added.to Enclosure 4.1 to- i log the first stage shell temperature before initial pressurizing.
< 10) New Step 2.10.14.8 was added.to. Enclosure 4.1 to verify that valves 2SM-25 and 33 open af ter , ].
l the "Off" pushbutton is selected on the turbine y
" Chest /shell warming" selector. ;!
,. 11) New Step 2.10.16 was added to Enclosure 4.1 to i 1 open valves 2SM-21 and 29. These valves do not !
receive a signal to open with'the "Off" push- ,'
button selected.
i
- 12) NOTE following old Step 2.15.5 concerning when ,
to reset the bleed steam check valves was .!
deleted. !
i
- 13) NOTE following old Step 2.46 concernin0 gener- :
ator output voltage being slightly higher than the indicated voltage was deleted.
i j.
I 18_8
9:: ,
fI M 14) Steps 2.49.1.8 an'd 2.49.2.9 were modified to require the " sync" pushbutton to be depressed <
when manually closing the' generator breakers,
- 15) NOTE was added prior to Step 2.49.2.4 to inform-the operators that the " sync" pushbutton should=
- be depressed,when manually closing a. generator breaker.
- 16) With 3-Arc admission .new Steps 2.1.1.6.1, .
i 2.1.1.6.2, 2.2.1.5 and 2,2.1.6 of Enclosure 4.2 l were added to verify that valves 2SM-25 and ':3 1 manipulate with the opening and closing of their respective' control valu.
- 17) New Step 2.3.2 wa= added to Enclosure 4.4 to- 4 verify that 2SM-25 opens when CV3 fully closes .
(approximately 65% of. full load). .
Ch'anges #6, 7. 8, 10,.11, 16 and 17 are being made because of the' implementation of VN #CE-1516.
Change #13 is being made because of the implementa--
tion of NSM #CN-20040. Changes l', 2, 3, 4, 5,-9 and i 12 are being made to enhence the effectiveness of this procedure. Changes #14 and 15 were needed to
-inform and ensure that the " sync" pushbutton is >
depressed when manually closing a generator breaker.
The generator breakers will not close manually if both the " sync" pushbutton and "close" pushbutton i are not depressed at the-same time.
' I Modifications made~in VN #CE-1516 and NSM #CN-20040 were evaluated in their appropriate Safety Evalua-tion and found to not contain any unreviewed safety
- questions. Because the majority'of the changes made in this retype resulting from the implementation of the previously named NSM and VN,' these-changes do i not contain any unreviewed safety ~ questions.
Changes 1, 2, 3, 4,-5, 9, 12, 14 and 15 are being made to improve the ease and effectiveness of this procedure and do not create any unreviewed safety ,
questions.
Since no functional changes are being-made to any safety systems and there are no relevant parameters in the bases to any Technical Specification, the margin of safety as defined in the bases to any Technical Specification is not reduced.
EP/1/A/5000/2E3 and High Containment Radiation Level: The only change EP/2/A/5000/2E3, made to this procedure was to modify the Annulus Retype #3 pressure setpoint given on Enclosure 1 Step #1.c and
- 1.d from -1.0 IN WC to -1.5 IN WC, This setpoint was changed by Design Engineering under exempt 189 4
l change CE-2466. This change will reflect the as built condition of the VE system. See exempt change CE-2466 for the 10CFR50.59 evaluation concerning this change. Per discussion with Mark Costello of Design Engineering, Design will pursue changing the above mentioned setpoint in the FSAR as listed under Section 6.2.3.3.
This change does not affect the procedure in a significant manner or alter its intent. Therefore, the above change does not represent an unreviewed safety question.
EP/2/A/5000/2E1, High Containment Pressure: The following_ changes Retype #3 are included in retype #3:
- 1) Modified Step #11.d.1) RNO to refer to Enclo-sure 5 for securing the Ice Condenser air handling units. Deleted wording to refer to OP/0/A/6200/08. Condenser Refrigeration System. References to this OP were deleted since this OP requires entry to Containment to complete the applicable enclosure. Created Enclosure 5 to accomplish the intent of this step, which is to turn off the AHUs, This change simplifies the procedure and makes it easier to carry out the required action.
- 2) Modified valve locations on Enclosure 1 & 2 and relocated the listing of various valves on these enclosures to reflect proper valve locations.
- 3) Modified the Annulus pressure setpoint given on Enclosure 4 Step #1.c and #1.d from -1.0 IN WC to -1.5 IN WC. This setpoint was changed by Design Engineering under exempt change _ CE-2467.
This change will reflect the' as built condition of the VE system. See exempt change CE-2467 for the 10CFR50.59 evaluation concerning this change. Per discussion with Mark Costello of Design Engineering, Design will pursue changing the above mentioned setpoint in the FSAR as listed under Section 6.2.3.3.
- 4) Corrected various typographical and format errors to ensure compliance with the EP Writ-er's Guide.
None of the above listed changes affects the proce-dure in a significant manner, or alters its intent.
Therefore, the above changes do not represent an unreviewed safety question.
190
qw ,
s EP/1/A/5000/2E1,. High Containment Presr.ure: ~The following changes Retype #7 are included in retype #7:
- 1) Modified Step #11.d.1).RNO to refer to Enclo- i sure 5 for securing the Ice Condenser air ;
handling units. Deleted wording to refer to OP/0/A/6200/08. Condenser Refrigeration System. References to this OP were deleted ,
since this OP requires entry to Containment to complete the applicable enclosure. Created Enclosure 5 to accomplish the intent of this step, which is to turn off the-AHUs. This change simplifies the procedure and makes it easier to carry out the required action.
- 2) Modified valve locations on Enclosure 1 & 2 and relocated the listing of various valves on these enclosures to reflect proper valve locations.
- 3) Modified the Annulus pressure setpoint given on Enclosure 4 Step #1.c and #1.d from -1.0 IN WC to -1.5 IN WC. This setpoint was changed by Design Engineering under exempt change-CE-2466.
This change will reflect the as built condition of the VE system. See exempt change CE-2466 for the 10CFR50.59 evaluation concerning this ;;
Per discussion with Mark Costello of change.
Design Engineering, Design will pursue changing a the above rentioned setpoint in the FSAR as ;
listed under Section 6.2.3.3.
- 4) Corrected various typographical and format-errors _to ensure compliance with the EP Writ-er's Guide.
None of the above listed changes affects the proce- -!~
dure in a significant manner, or alters its intent.
Therefore, the above changes do not represent an unreviewed safety question.
EP/2/A/5000/1C, High Energy Line Break Inside Containment: -The Retype #5 following changes are included-in retype #5:
- 1) Modified the Annulus pressure setpoint given in Step #17.b and on Enclosure 4 Steps #1.c and
- 1.d. from -1.0 IN WC to -1.5 IN WC. This setpoint was changed by Design Engineering under exempt change CE-2467. This change will reflect the as built condition of the VE ,
system. See exempt change CE-2467 for the 10CFR50.59 evaluation concerning this change.
Per discussion with Mark Costello of Design Engineering, Design will pursue changing the 191
n , 1 g
i
-above mentioned setpoint in the FSAR as listed.
under Section 6'2.3.3.
- 2) -Added Step #16 which dispatches an operator _to close 2NI-208 and renumbered all-subsequent steps. Closing valve 2NI-208 was suggested by the safety evaluation for NSM CN-20442 which ;
installed 2NI-208. The reason given for-closing this valve is to reduce post-LOCA ,
radiation levels'in the Auxiliary Building .
during sump recirculation. .The problem of_high radiation levels is mainly a concern following a large break LOCA with significant fuel failures when sump recirculation is initiated.
This_ step was previously in EP/2/A/5000/02, Reactor Trip or Safety Injection. This step is being moved to this procedure so that it will only be necessary to close 2NI-208 following a large break LOCA. This accomplishes the intent of the step without having to close the valve -i during other accidents when it is unnecessary.
- 3) Changed wording of Item D on Enclosure 1 from " [
... REFER TO Step 15 ..." to "... REFER TO '
Steps 15 and 16 ..." This was done to in 4- response to change #2 listed above. ,
- 4) Corrected various format errors to ensure compliance with the EP Writer's Guide.
None of the above listed changes affects the proce-dure in a significant manner or alters its intent.
Therefore, the above changes do not represent an I
unreviewed safety question.
EP/1/A/5000/1C, High Energy Line Break Inside Containment: The Retype #9 following changes are included in retype #9: ;
- 1) Modified the Annulus pressure setpoint given in '
-Step #16.b and on Enclosure 4 Steps #1.c. and
- 1.d. from -1.0 IN WC to -1.5 IN WC, This setpoint was changed by Design Engineering under exempt change CE-2466. This change will reflect the a3 built condition of the VE system. See exempt change CE-2466 for the 10CFR50.59 evaluation concerning this change.
Per discussion with Mark Costallo of Design Engineering, Design will pursue changing the above mentioned.setpoint in the FSAR as listed under Section 6.2.3.3.
- 2) Corrected various typographical and format errors to ensure compliance with the EP Writ-er's Guide.
1 12
n None of the above listed changes affects the proce-dure in.a significant manner or alters its intent.
Therefore, the above changes do not represent an unreviewed safety question.
MP/0/A/7400/20 Diesel Engine Fuel Oil Filter and Strainer Removal and Replacement: This is a previously approved and fully reviewed procedure that is being upgraded to the new procedure format. The changes are a result of the new format, are minor in nature, and are NOT.
technical. The minor changes do NOT create an unreviewed safety' question, and have'no effect on referenced sections. -
/
A new section has been added; Section 11.6'. Section 11.6 provides a method for removal and replacement i'l
'.. of the fuel oil strainer housing. The incorporation of, is to consolidate inspections and activities i under one procedure for this component. This <
addition does NOT create an unreviewed safety ;
question and has no effect on identified FSAR l sections. j l
l l
l 1
)
i I Y 1
19_3
-