ML20029B330
| ML20029B330 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/31/1990 |
| From: | Creel G BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9103070065 | |
| Download: ML20029B330 (24) | |
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8 B ALT I M O R L GAS AND ELECTRIC CHARI ES OENTER. P. O. BOX 1475. BALTIMORE, MARYLAND 21203 GroRot C. CREFL vict P.e sioe u u m t..t~i.., March 1,1991 ooo n e ..n U. S. Nuclear Regulatory Commission Washington,DC 20555 NITENTION: Document Control Desk SUlljECT: Calvert Cliffs Nuclear Power Plani Unit Nos.1 & 2; Docket Nor.. % 317 & $0 318 Reoort of Changes.Testr. and Exocriments
REFERENCE:
(a) 10 CFR 50, Paragraph 50.59(b)
Gentlemen:
As required by the above reference, please find enclosed our annual report of changes, tests, and experiments completed on Calvert Cliffs Unit 1 and/or 2 under the provisions of 10 CFR 50.59(a),
including a summary of the safety evaluation for each, This report covers the period from Ja'>uary 1,1990 through December 31,1990.
Items in the report are referred to by Facility Change Request (FCR), Field Engineering Change (FEC), Temporary Modification or Miscellaneous Activity number, Should you have any questions regarding the contents of this repoo, we will be pleased to discuss them with you.
a Very truly yotis, ,.. ,
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GCC/J Bil/BSM!<"m - r 3
Enclosure 1: Annual Report of Changes, Tests and Experiments (20 pages) v eA 9103070063 901231 8
8 Document Control Desk March ),1991' i Page 2 cc: D. A. Brune, Esquire J. E Silberg, Esquire R. A. Capra, NRC D. O. Mcdonald, Jr., NRC T. T. Martin, NRC L E Nicholson, NRC R.1. McLean, DNR J. II. Walter, PSC e
1 9
89 h
Document Control Desk March 1,1991 '
Page 3 bec: 4 A Tiernan/A.J.Slusark
'. Micrnickl W. R. Corcoran/F, J. Munno C.11. Cruse /P. E. Katz R. C. DeYoung
. R. M. Douglass/R. F. Ash R. P. Heibel/T. N. Pritchett C. P. Johnson i C. C, Lawrence, lil/A. R. Thornton
. R B. Pond, Jr./S. R. Buxbaum L B. Russell /lL E. Denton/J. R. Lemons W. A. Thornton/E.1. Bauereis G. L. Adams (2)
A. B. Anuje J E. Baum J. J. Connolly G. L Detter G. J. Falibata
! D. Graber .
, D.V.Graf
' R. E. Nagel B. S. Montgomery R. C. L Olson P. A.Pieringer Technical Library M. J. Warren -
L O. Wenger J. B. Borowski t _
ENCLOSURE (1)
ANNUAL REPORT OF CHANGES, TESTS AND EXPERIMENTS L
l CALVERT CLIFFS NUCLEAR POWER PLANT l 1990 l -.
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12CFR50.59 Annual Report Pace 1 FCR 82-79 FCR 82-79 moved power cables for #11 Containment Air Cooler, but did not reflect the change on FSAR Figure 8-3. This Safety Evaluation was performed to incorporate the as-built change from penetration 1ZEB1 to 1ZEA4 for #11 Containment Air Cooler on FSAR Figure 8-3. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.
FCR 84-1086 This change replaced the existing unqualified thermocouple assembly, signal converter and temperature indicator for Unit 2 Containment Dome Temperature with environmentally qualified units in accordance with Reg. Guide 1.97, Category 2 requirements. The Safety Evaluation concluded there was no unreviewed shfety question or a change in the Technical Specificwtions.
FCR 85-1022 Sucolement No. 15 This change removed the tubing and piping that wac abandoned in place after removal of the original screen wash system. In addition, two cables that were abandoned in place will be retagged as spares. This activity does not affect any safety related equipment. However, because it is a change to FSAR figures 9-23-2 and 9-28A, it required a Safety Evaluation. The Safety Evaluation concluded there was no unreviewed safety question or a change in the Technical Specifications.
ECR 87-69 Sucolement 2 This activity is necessary to correctly identify the containment isolation valves for penetration Nos. 9 and 10 and to bring the FSAR into agreement with the Technical Specifications.
This activity revi, ' FSAR Figure 5-10 Sheet 8 to correctly identify the contais, int isolation valves for penetration Nos. 9 and 10 and to furthe 71arify this in Section S.2.2. This Figure
-identified valves SI- e6, CV-4151, and CV-4160 for penetration No. 9 and SI-316, CV-4150, and CV-4159 for penetration No. 10 as being the respective penetration's containment isolation valves.
Also, Section 5.2.2 will be revised to further clarify that it does not apply to penetrations Nos. 9 and 10 since the containment spray system is an accident mitigation system which will not be i=olated post-accident until containment pressure is reduced or a failure of.the spray system occurs. Per Te,hnical Specification Table.3.6-1 and the following discussion, the correct isolation valve numbers for penetration No. 9 are SI-326 and SI-340 and for penetration No. .10 are SI-316 and SI-330.
During normal plant operation, check valves SI-326 and SI-340 for Penetration 9 and SI-316 and SI-330 for Penetration 10 remain closed as there is no flow in the containment spray lines. For both Penetrations 9 and 10, the isolation valves for tne spray
. l 10CFR50.59 Annnal Report Pace 2 l header (CV-4150, CV-4151) and the charcoal spray system (CV-4159, CV-4160) are both located downstream of check valves SI-340 for Penetration No. 9 and SI-330 for Penetration No. 10 and are normally closed. Upon receipt of a Safety Injection Actuation Signal (SIAS), valves CV-4150 and CV-4151 open to allow flow to the spray ring nozzles. These two valves also fail open upon loss of power or air. In addition, valves CV-4159 and CV-4160 are opened by a handswitch and also fail open on a loss of power or air. With CV-4150 and CV-4151 op2n after a SIAS, the check valves provide containment isolation since the spray headers are open to the containment atmosphere. Technical Specification Table 3.6-1 correctly lists SI-326 and SI-340 for Penetration No.
9 and SI-316 and SI-330 for Penetration No. 10 as the containment isolation valves. FSAR Figure 5-10 Sheet 8 and Section 5.2.2 will be revised to reficct this. The Safety Evaluation concluded there was no unroviewed safety question or change in the Technical Specifications.
FCR 87-129 This change repowered the 72 foot computer room air conditioning units such that a minimum of one unit will be operable during a Loss of Offsite Power (LOOP) . This el:nge included new power cable installation from the computer room to s itch gear rooms.
The reason for the change is because the Safety Parameter Display System (SPDS) must be available during normal and abnormal events including LOOP as required by NUREG-0737 Supplement 1. _The Safety Evaluation concluded there was no unroviewed safety question or a change in the Technical Specifications.
FCR 88-11 Suco.5 As described in the FSAR, the original chlorination design, was
! to inject a chlorine solution into the intake weir just I
downstream of the traveling screens. The chlorine concentration was cont??alled at this point rather than at the plant effluent as required by the NPDES permit.
In order to meet the requirements of the NPDES permit, it was decided in 1976 to redirect the chlorination system to serve only the Salt Water (SW) system. Although the physical changes to the plant have been made, the one paragraph description of the Chlorination system in the'FSAR has not been revised to ref]ect these changes. Additionally, after reviewing the safety analyses
, in FCR 76-99 and associated FEC's, although adequate for that time' period, it is prudent to revisit the changes made under FCR 76-99. Therefore, this 602 5L evaluation was written to address the overall differencet between the FSAR description and the as-built Chlorination sys' ass The Safety Evaluation concluded there was no unreviewed safeti question or a_ change in the Technical
, Specifications.
10CFR30.59 Annual ReDort Pace 3 FCR 88-031 This change replaced the Durametallic mechanical seal on Low Pressure Safety Injection (LPSI) pumps 11, 12, 21, and 22 with a Borg Warner cartridge type seal assembly due to unacceptable amounts of leakage past the shaft. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications, FCR 88-3000-1 During the refueling outage following completion of Unit 1 cycle 9, all the reload fuel for cycle 10 operation were examined. As a result, four fuel pins were removed and replaced with stainless steel dummy rods. The Safety Evaluation concluded there wac no unr9. viewed safety question or a change in the Technical Specifications.
FCR 88-3002 The previous maximum enrichment limit for the fuel handling equipment (fuel upender, fuel inspection elevator, and new fuel elevator) was 4.1- wt% U-235. The fuel handling equipment was upgraded to 5.0 wt4 0-235 to enable movement of higher enriched fuel. assemblies between the spent fuel storage racks and the reactor core, and the now fuel storage racks and the spent fuel pool. The Safety Evaluation concluded there was no unroviewed safety question or a change in the Technical Specifications.
EC.8_jj-089 Sucolement ',
Boric Acid deposits were discovered at several pressurizer heater penetrations on Unit 2. It was established through destructive and-non-destructive examinations that these sleeves were leaking as a result of Primary Water Stress Corrosion Cracking (PWSCC).
This change is to replace Unit 2 Pressurizar Heater Sleeves-except at penetration H-3. The Safety Evaluation concluded there was no unroviewed safety question or change in the Technical Specifications.
FCR 89-089 Suonlement 6 This change plugged Unit 2 Pressurizer heater penetration H-3.
This evaluation only addrenses the mechanical engineering aspects of this change. Boric acid deposits were discovered at several
. penetrations. A large diameter core sample was bored out from the pressurizer at penetration H-3 to include the weld, and deutructively examine it for root'cause analysis. Heater reinstallation *tay encounter alignment problems, therefore it was decided to plug the penetration. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technica1' Specifications.
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10CPR50.59 Annual ReDort Pace 4 J
FCR 89-089 Supplement 7 i
This. modification changed the power supply to the pressurizer heater at location C3 from Non-1E MCC 210PH to a diesel bam.ad
'(1E) power suptt, to the heater at location H3 and reconnecting it.to the heater at-location C3.
This change was necessary as pressurizer heater penetration.H3 was permanently plugged during the Unit'2 pressurizer repairs.
In-order to maintain a redundant 450 kw heater capacity powered from.a 1E power supply (reference letter from R. Reid - NRC'to A.
Lundvall - BG&E dated April.-7, 1980) after the-present 1E heater location H3 is permanently plugged, the pressurizer heater in location C3 must be powered from MCC 211PH, instead of MCC 210PH
- which'is non-1E. _This change will not place any additional-loading on the diesel generator as heater C3 requires 12.5 kw which were the same power requiremente as heater H3. The Safety Evaluation concludes there was no unreviewed safety' question or
- change in the Technical Specifications.
FCR 89-98 ThisLchange was made to replace the Unit 1 and.2 Safety. Injection
- Tank "non-code" relief valves with new Crosby.Model JOS ASME Section VIII relief valves. 'The FSAR states that the SITS were-designed, constructed, and have overpressure protection in accordance with the ASME Code,Section III, Class C.
g Additionally, the nitrogen relieving capacity of the' safety E Injection Tank relief valves, were insufficient if the nitrogen L supply line control valve would-fail in the open position. The new-Section VIII relief valves satisfy the FSAR code
' requirementc. The Safety Evaluation concluded there was no unreviewed safety question or. change in the-Technical
- Specifications..
L FCR 89-0171 L
ThisJmodification Will eliminate an uu;' anted'shutdownfof Diesel I Generator #11 from a short circuit caused by a postulated control L ~
room ~ fire. This is in agreement with~the Diesel Generator #12 and Diesel GeneratorE#21 stop-circuits-isolation feature initiated by FCR 81-1052. . In: addition, the requirements of AOP-E 9Ac(Control Room Evacuation and Safe-Shutdown due to a Severe-Control Room Fire) are:also_being met.- ,The Safety Evaluation-concluded there!was no unreviewed- safety question or a change:in' the~ Technical Specificatic:w.
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10CPR50-59 Annual Report
. Pace 5 FCR 89-180 Supplement 2 Many Saltwater Control Valves must remain fully operational post-LOCI. During performance of an Engineering Test Procedure, it was discovered that many of the air operated control valves which utilize Safety Related (SR) air accumulators would not have performed as expected after a loss of normal Non-Safety Related (HSR) air supply. Of specific concern are the valves which are inaccessible post-LOCI and are required to align the Saltwater System to the overboard mode of operation upon a rupture of a Saltwater Discharge line after a Recirculation Actuation Signal (RAS) occurs. This change upgrades electrical circuits, supplies SR air to control valves, provides a cross tie between SR and NSR l air, and modifies air tubing. The Safety Evaluation concluded there was no unreviewed safety question or a change in the i
- Technical Specifications. l 1
FCR 89-139 l A 1977 LTOp commitment stated that both PT-103 and PT-103-1 1 (pressurizer pressure transmitters) would provide a computer generated high pressure alarm and type written printout. This
- modification allowed PT-103 and PT-103-1 to provide the required alarm and printout function. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.
FCR 89-3001 This change removed the following sentence from the UFSAR, "At no time during the transfer from the reactor core to the spent fuel storage rack is there less-than 112 inches of water above e fuel assembly." This sentence was incorrect, had no basis and unnecessarily restricted fuel-assembly movement. The Safety Evaluation concluded there was no unreviewed safety question or a
. change in'the Technical Specifications.
Egg 90-10 The text of the.UFSAR,Lin a number of cases, is too vague to !
adequately define separation of cables and raceways in redundant scoaration groups. This evalvation was written to modify Section .'
-8.6 in the UFSAR, allowing-the use et evaluations when determining acceptable sepa*.ation. Tha probability of failure will not increase by evaluating the actual separation required.
The Safety Evaluation concluded there was nt unrevieaod safety quer *'n or change in the-mechnical Specifications.
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10CFR50.59 Annual red 9rt Pagg_1 FCR 90-010 Sunclement 1 The FSAR implies that only silicate separation barriers are used for radeway/ cable tray separation in the plant. This is not the case as sheet metal tray covers are used also. This change will clarify the FSAR, to specify which materials may be used as separation barriers. The Safety Evaluation concluded there was no unroviewed safety question or e change in the Technical Specifications.
FCR 90-011 The Component Cooling Watur (CCW) flow-rate +- the Shutdown Cooling Heat Exchanger (T,DCHX) was reduced du.Ang all modes of operation, due to previous manual valve position causing heat exchanger tube rattling during normal shutdown cooling operation.
Performance Enginec ring determined that at shell side flows of 3000 gpm, considerable tabe to tube and tube to baffle contact occurs, and increases in frequency and intensity as flow increases. Engineering Test procedure 89-77 set the 11 and 12 i SDCHX inlet manual valves to provide a flow of 2500 gpm to each l heat exchanger uhen two CCW pumps are operating and all major CCW j londr 'CW flow' thro"gh the thell side of the SDCHx) are secured.
(cona Sent, ev s., rator, 1etdown). With this position hold, the CCd mystera was aligned for normal and accident operation, Given a LOCA with a LOOP and one OG failure, the flow through each SDCHx is 1800 gpm and this flow rate was used in the LOCA analysis.
As a tenult of the change in CCW flow to the SDCHxs, two calculations were performed that affect-information in the FSAR.
- 1. LOCA The LOCA analysis was reviewed to determine the post-RAS containment response given a CCW fl.ow of 1800 gpm.
The results show that-it will take a longer period-of time for containment to be brought back to 120 F (<30 days). '
The EQ group of EAU was able to demonstrate acceptable qualification for a 30-day cool-down time period (see EQDR 14, Rev. 1).
- 2. Normal Cool-down The FSAR states that the RCS can be brought from 300 F to refueling temperature in 27 1/2 hours after shutdown. Because of the reduced CCW flow, the cool-down extends to 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
The Safety Evaluation concluded there was no 'rireviewed safety question or change in the Technical Specifications.
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,10CFR59 59' Annual ReDort Pace 7 FCR 90-21 Prior to 1986,.the-manufacturer's technical manual did not specify system accuracy for the Delphi Model K-IV Hydrogen Analyzers. Then in-1986, the. manufacturer's representative established a system accuracy of 5 percent of full scale-(0.5 percent H2)-which was then documented in a revised technical manual (12-277-24). During surveillance testing of the Hydrogen
- Analyzers, required by Technic:1 Opecifications 4.6.5.1 and 4.6.5.2i maintenance could not achieve the recorder or indicator loopz accuracy stated in BG&E March 14, 1983 letter to the NEC.
- Therefor, calculation 190-16.was performed by I&C Engineering to determine the total loop, uncertainty (accuracy) associated with the recorders and indicators of the hydrogen analyzers using present day methodology.
The methodology used in this calculation is conservative in that little use of square root sum of the squares (SRSS) is used. A l margin above STP tolerance is allowed without approaching the calculated loop uncertainty.
- Loop Uncertainty for Indicator Loop = 10.26 % (1.0% H2)
Loop Uncertainty for Recorder Loop = 9.40 %-(0.94% H2)
The Safety Evaluation concluded there was no unreviewed safety _'
- question.or change in the Technical Specifications.
FCR 90-036 The_ previously installed Unit 1 and 2 SG blowdown tank RVs had
-insufficientLcapacity. This change replaced the SG blowdown tant RVs with new Crosby style JBS RVs to comply with:Section VIII of
- the ASME code. The Safety Evaluation concluded there was no unreviewed-safety' question or change in the Technical
- Specifications, i JCR 90-64
' This change enhanced the reliability of the steam driven Auxiliary Fcadwater System by allowing the turbine governor to
- accelerate the AFW pump in a more controlled manner in order ~to eliminate turbine feed pump trip due to' initial over-speed. The change modifies the control schemes of the AFW Main Steam -
, Admission Valves.(MSAV) 1-CV-4070.and 1-CV-4071. The. Safety
-Evaluationtconcluded there was no unreviewed safety question or
- change in tha Technical Specifications.
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10CFRip.59 Annual Reoqxt Pace _H FCR 90-68 This activity accepted the "as-found" plant condition for the loose Calvert Cliffs Unit 2 core support barrel (CSB) snubber H pins.- During the 1989 outage, a visual examinat10n was performed on the core stabilizing lugs. This examination revealed that two of the snubber pins were partially-sticking out, The first pin is located in the snubber assembly at the 120 location and was sticking out approximately 3/8". BC&E attempted to pull the pin out using a robot submarine. The submarino accidentally bumped ,
into the pin, knocking the pin back into the hole, The pin is l l now in its proper-location. A cecond questionable pin is located L in the snubber assembly at the 180 0 orientation. It appears as l if this pin is slightly out of its proper position. There is a possibility that-the small locking pins located in the CSB snubbers in the reactor vessel downcomer may loosen and fall into the flow-stream. This-"as-found" plant configuration introduces concerns with loose particles in the reactor coolant system
'(should the pin come completely out) and increases the L possibility that a shim retaining bolt can loosen and back out.
l In' conclusion the Safety-Evaluation-showed that-the probability e of-. occurrence or.the consequences of an accident or malfunction I of-equipment important-~to' safety previously evaluated in the FSAR is not increased. The Safety Evaluation also concluded there was no unreviewed safety question or change in the Technical
-Specifications.
FCR 90-87 L
L This change revises FSAR Section 9.4 to.allew the four removable L Spent Fuel Pool (SFP) spool pieces to remain installed in the.SFP L system during all modes of' operation. FSAR Section 9.4 refers to
! the. spool pieces:as temporary tie-ins to be used as a means of t
augmenting the heat removal capacity to_SFP cooling when 25/3 coras are in the pool. The design intent for the temporary spool pieces,was to. keep the systems _ separate while. allowing for-hook-l up for:a very rare occurrence.of cooling augmentation. This P independence can.be satisfactorily achieved via isolation. valves l without having to remove the spool pieces, thereby eliminating a u valid ALARA concern. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical
' Specifications.
_FCR 90-095 This activity resolved minor inconsistencies between the defined
- quality standards for field erected tanks as described in FSAR Section 6.3.5.1 and the quality standards imposed in the purchase 1
specification for the field erected tanks. Review of FSAR Section-6.3.5.1, applicable-P& ids, related documents and applicable specifications indicated the change to the FSAR was necessary and no other hardware problems exist. The Safety Evaluation concluded there'was no unreviewed safety question or a change in the Technical Specifications.
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-10CFR5Qt19_ Annual Reoort Pace 9 FCR 90-146 This change-was necessary to show in the FSAR the corrected postulated dose-rates from Units 1 and 2 Emergency Personnel Air l Locks and to include the corresponding information for the Units '
l'and 2 equipment hatches. The Safety Evaluation concluded there was no unreviewed safety question or a change in the Technical Specifications.
FEC 85-1041 l This modification removed differential pressure indicator 2-PDI- 1 2420 including all of the associated process tubing and supports for the-indicator. This was necessary as the dif'erential
. pressure- indicator and process tubing interfered with the
-installation of Main Steam line support 20" EB-3-2002-R14. These components are no' longer in service and removing tnem will facilitate the installation of this main steam line support. The ~
Safety Evaluation concluded there was no unreviewed safety question or a change in the Technical Specifications.
FEC 87-0074-011 This change rerouted Unit 2 Reactor Coolant Pump vapor seal 1:aK off lines from_the Reactor Coolant Drain Tank (RCDT) to the
-containment sump. This prevents spray at the seal due to back pressure at the: tank. The Safety Evaluation concluded there was nosunreviewed' safety question or a change in the Technical Specifications.
FEC 89-01-149 iThis; change replaced-the existing shaft packing seal with a mechanical seal in Condensate, Pump No.-11. Minor piping modifications-were<also.done to accommodate this change. .-The Safety [ Evaluation' concluded there was no__unreviewed safety-question:or a change in the1 Technical Specifications.
'FEC 89-01-473
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~This change install'ed~ drain lines off existing drain taps on all Emergency Core Cooling System 4(ECCS) basket-strainers. This will
-allow draining of the-strainers to the: Saltwater system instead of into-floor drains, thereby lessening the amount of liquid waste. -The_ Safety Evaluation concluded there was no unreviewed safety question'or a change in the Technical Specifications.
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ADCFR50.59_ Annual _Pgoort Page 10 PEC 89-03-35 This activity approved the change of Salt Water pump component materials. This was requested by the vendor due to market availability. Some of the components involved included, impeller shaft, impeller shaft nut, impeller shaft sleeve and thrust bearing shaft ring. This activity found the substitute materials acceptable. The Safety Evaluation concluded there was no unreviewed safety question or a change in the Technical Specifications.
EEC 89 ,98-1 and 2 This activity relocated the safety relief valves 1(2)-RV-211, 221, 231, and 241 for all the Safety Injection Tanks (SITS) on Units 1 and 2. This was required because of the failure of' seal welds at pipe connections for the SIT vent lines upstream of 1-RV-221 and 241. The relief valves were relocated from the SITS vent lines to the nitrogen fill lines upstream of the SIT, near the base of the tanit. This will reduce the local stresses on the vent lines fo; the SITa where the relief valves were attached.
The Safety Evaluation concluded there was no unreviewed safety i question or a change in the Technical r occifications. l l
FEC 90-01-17 This change replaces the body of 2-CV-5209 and the two flanged spool pieces in the 30" piping downstream of the service water heat exchanger salt water discharge throttle valve, 2-CV-5210, with a single, robber lined spool piece.
The salt water svstem was sized for a loss of coolant accident.
During normal or eration, the flow through the service water heat exchanger is tLrottled at the heat exchanger discharge centrol valves 2-CV-5210 and 2-CV-5212. The turbulent flow associated with throttling these butterfly valves caused erosion of the valvGs and the downstream piping. To mitigate this effect, orifice valves, 2-CV-5209 and 2-CV-5214, were installed downstream of the throttle valves (FCR 75-1100). These valves could be opened to provide full flow during accident conditions and cicsed during normal operations to move the turbulence away from the discharge control valves. Unfortunately, erosion was even more severe after installation of the new valves.
FCR 80-0017 removed the internals and actuators of 2-CV-5209 and 2-CV-5214 and inatalled rubber lined spool pieces downstream of the discharge control valves. FCR 82-1012 replaced valves 2-CV-5210 and 2-CV-5214 with rubber lined valves. These modifica' ions appear to have resolved the localized pipe erosion problem, ine Safety Evaluation concluded there was no unreviewed safety question or change in th7 Technical Specifications.
10CFR50.59 Annual Renort Pace _11 FEC 90-01-028 While working on one t the Main Feed Regulating Valves (MFRV) it was determined that some washout of the valve body had occurred.
This washout was rrpaired by a weld buildup. A radiographic examination (RT) el the repair weld was required but was not possible to perform on the entire weld build-up due to the non-uniform valve geametry. This Safety Evaluation was performed so that a magnetic particle (MT) examination could be done instead.
The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications.,
FEC 90-01-143 The installed Unit 1 and 2 SG blow-down tank R.V.s had insufficient capacity to satisfy section VIII of the ASME code The procurement and installation of a new SG blow-down tank R.V.
could not be accomplished in a timely manner. BG&E opted for a !
temporary fix, installing flow limiting orifices in the SC blow-down lines to limit the mass flow rate to the blow-down tank.
Limiting-the mass flow-rate to the tank will ensure the blow-down tank will not become overpressurized and will retain its structural integrity. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical.
Specifications.
FEC 90-01-173 A
Installation of isolation valves upstream and downstream of the boundary check valve 2-IA-310 has been completed. Also, a pressure point consists of teeing off of the main line with a7 isolation valve and a drain. This change was made because it is required to periodically test all safety-related and non-safety related boundary check valves. By adding these isolation valves and pressure points the check valve 2-IA-310 can be tested with minimum impact to the plant. The Safety Evaluation concluded thete was no unreviewed safety question or change in the Technical Specifications.
FEC 90-01-132 Logic diagrams 61058A (1LD58A) and 63058A (2LD58A) were revised to more accurately reflect SIAS and RAS operation for Component Cooling and Service Water Heat Exchancers' salt water valves 1(2)
CV-5160, 5163, 52G6, 5208, 5210,'5212. These revisions to the logic diagrams reflect information from the text of FSAR Section 1 9.5 and existing schematics. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications, i
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10CFR50.59 Annual-Report Pace 12 FEC 90-01-238 This activity added: low = point drain lines to the Unit 2 feedwaterL recirculation lines, upstream of the condenser. These low point drains are necessary to allow for draining and t' lushing of the feedwater piping which the condenser is unavailable and d '
lated.
The Safety Evaluation cor.luded there was no unreviewed nafety question or a change in the Technical Specifications.
FEC 90-01-250 Auxiliary. Boiler Steam system steam trap 0-ST-1855 was modified from two-inches to-one inch-size due to excessive, unnecessary loss of steam, reducing system.officiency. The Auxiliary Boiler
-Steam system is a non-safety related system. However, this FEC requires n' change;to FSAR figure 10-6 and therefore, this Safety Evaluation was performed. The Safety Evaluation concluded there was no-unreviewed safety. question or a-change in the Technical Specifications. ,
-FEC 90-01-265 This: change' changes"BG&E drawing 60-227-E to reflect manual drain and ventLyalves 1-BD-187, 188, 189, 190, and 191 and isolation valves 1-BD-192 and 193 in UnitL1 Steam Generator Blowdown and Recovery System piping. All-of the valves are-in Non-Safety Related piping and not_ described in_the t xt of the FSAR.
However, this Safety Evaluation is requi:.vd becausp 60-227-E is shown'in the FSAR.='.The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications..
FEC'90-1-298 ,
This FEC proposed the replacement of the: original "G" type buffer spring-with'a-vendor 1 recommended = stiffer "F"-type buffer springs,
-in the: Avk12 iary :Feedwater (AFW)- system turbine governor compensacory anit'; _The replacement buffer. springs are proposed for1both-~ Unit- 1 ~and 2 AFW system turbine governors. _The "F" type buffer =upring kill ~ help dampen 1the sudden minor-load changes experienced 1by-che governor, but allow it*to respond-to major load changes to govern 1the turbine speed.- The Safety Evaluation <i concluded there was no unreviewed safety question or a-change in the Technical = Specifications.
LFEC 90-01-406 This activity involves updating Unit 1 elevation 12 feet Turbine Building ' location drawing (FSAR . Figure 1-20) nto reflect the as-built' location. of-the Mechanical Chiller for the Turbine plant samples 1and-removal of the chemistry lab.. .This activity is-non--
safety related and does not affect any safety'related-equipment,.
butLa Safety Evaluation in necessary since the. activity requires a' revision to FSAR figure 1-20.- The-Safety-Evaluation. concluded-
10CFR50.59 Annual RoDort Pace 13 there was no unreviewed safety question or a change in the Technical Specifications.
FEC 90-01-491 This change removed and reinstalled the air components on the Unit i feedwater regulating control valves. The air supply comporients were retubed using flexible hose installed between the control valve and solenoid valves which are mounted on a support frame. The reason for this activity is due to two (2) incidences where a less of instrument air was due to shear and fatigue caused by the vibration of the feedwater regulating valves. The Safety Evaluation concluded there was no unreviewed safety question cr change in the Technical Specifications.
FEC 90-01-537 This change replaces the body of 1-CV-5209 and the flanged spool piece in the 30 inch piping immedit,elv upstream of 1-CV-5209 with a single, rubber lined spool pv.~-
The salt water syster .e.c cized for a loss of coolant accident.
During normal operation, the flow thr.ough the service water heat exchangers is throttled at the heat exchanger discharge contrul valves 1-CV-5210 and 1-CV-5212. The turbulent flow associated with throttling these butterfly valves caused erosion of the valves and the downstream piping , To mitigate this effect, orificed valves, 1-CV-5209 and 1-CV5214, were installed downstream of the throttle valves.(FCR 75-1100). These valves could be opened to provide full flow during accident conditions o _and clcse;l duri ng normal operat i ons to move the turbulence away from the discharge control valves. -Unfortunately, erosion was even more severe after installation of the new valves.
FCR 80-0017 removed the internals and actuators of 1-CV-5209 and 1-CV-5214 and installed-rubber lined spool = pieces downstream of ,
.the discharge control valves. FCR 82-1012 replaced valves 1-CV-5210 and 1-CV-o212 with rubber lined valves. These modifications appear to have resolved the localized pipe crosion problem.
The. flange on the second spool piece between 1-CV-5210 and the body of 1-CV-5209 was leaking. Replacing this spool piece and the body of 1-CV-5209 with a single spool piece stopped the leakage and returned the piping run closer to its original configuration. A rubber. lined spool piece was used to mitigate the erosion effects associated with 1-CV-5210 and to provide good corrosion resistance. The Safety Evaluation concluded there was nc unreviewed safety question or change in the Technical Specifications.
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10CFR50.59 Annual Report ___ Pace 1_4 FEC 90-1-578 This change will delete level transmitters 2-LT-311A, 321A, 331A, and 341A from FSAR figure 6-10A. These level transmitters were never installed because of problems associated with purchasing level transmitters compatible with the Safety Injection Tanks existing instrument taps. The Safety Injection system, as described in the FSAR is not affected. The Safety Evaluation concluded tnere was no unreviewed safety question or a change in the Technical Specifications.
FEC 90-01-579 This change involves updati'ig the Unit 2 Instrument Air System piping and Instrument Diagram (P&ID) to show the as built location of the Instrument air headers which supply MSR Drain Tank 23 and First Stage Roheater Drain Tank 23 pneumatic level centrol instruments. A safety Evaluation was necessary since the change creates a revision to P&ID M-454 Sh. 2 which is FSAR Figure 9-28A. The Safety Eva.uation concluded there was no unreviewed safety question or a change in the Technical Specifications.
FEC 90-01-819 This change is to allow the design pressure of the feedwater piping, between the containment penetrations and Steam Generators to be reduced from 1500 psig to 1400 psig. The design pressure of 1500 psig for the feedwater piping between the containment penetrations and the Steam Generators is very conservative. This statement is based on the fact that the Steam Generators have pressure relief via the Main Steam Code Safety Relief Valves, which-have staggered tettirgs ranging from 1000 psig to 1G50 psig. The maximum Stear Generator pressure as found in Chapter 14 of the FSAR, is 1074 poig which occurs during an Asymmetric Steam Generator Fvent, Those are the worst case precsure transients for the feedwater pipe in containment. At normal operating conditions the Steam Generator sustained pressure is about 850 psig. This is the pressure that meets the code definition of Design Pressure, which is the maximum sustained pressure. The pressure drop as given in the original design calculations from the Steam Generator to the flow element is about 40 psi, for a pressure at the FE's of 8fs0 psi. Therefore, a rcduction in the Design Pressure to 1400 psig will still envelop the worst case, and design conditions, will not compromise the safety or design basis of tbs feedwater line.
However, this change will allow us to extend the useful service life of the feedwater flow elements which have been found to have wall thinning due to erosion / corrosion of the interior curfaces. The Safety Evaluation concluded there was no unreviewed. safety question or a change in the Technical Specificat1ons.
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'10CFR50.59 Annual Report Paos 15 Igap.grary Modification f 1-90-61 The previously installed Unit 1 and 2 SG blowdown tank RVs had insufficient capacity. The capacity requirements, which must be satisfied, are specified by Section VIII of the ASME Code.
l l The procurement and installation of a new SG blowdown tank RV could not be accomplished in time to support the April 1990 Unit 1 plant start-up schedule. Thus, BG&E opted for a temporary fix.
Flow limiting orifices (1-FO-4017,4018) were installed in the common SG blowdown lines via FEC 90-01-143. Additionally, CCI-117 Serial #1-90-26 was issued to chain shut the surface blowdown manual isolation valves (1-BD-194, 196, 198, 200). Both activities were implemented so as to limit _the maximum mass flowrate that could be supplied to the blowdown tank. Limiting the mass flowrate to the tank will ensure that the blowdown tank will not become overpressurized and will retain its structural integrity.
l Unit 1 was shut down for the 1990 Eddy Current Outage. During this outage, the steam generators were drained via the SG blowdown lines. The installed orifice plates overly restrict the L draining of_the steam generators. Therefore, the orifico plates l l were replaced with non-flow restricting spacer plates. !
The SG blowdown line orifices limit the mass flowrate to the SG blowdown tank so that the tank does not become overpressurised.
l Removing the orifices from the blowdown lines means that the
! blowdown tank will no longer have adequate overpressure protection. To ensure that the tank was not placed back in service without overpressure protection, this temporary modification was a Unit 1 mode restraining activity because it limited the-RCS temperature to less than 200 F. This temperature limit prohibited the plant from entering mode 4 and ensured that i the blowdown-tank pressure remained less than its 200 psig design l pressure. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specifications, l-Temporary Modification 1-90-13Q This temporary modification uas to suppress nuisance alarms on l
Channel A of the Unit 1 Reactor Vessel Level Monitoring System l
(RVIMS) resulting from an inoperable heated junction thermocouple (HJTC) on sensor number eight. The control room trcuble end lov level alarm remain in alarm. The temporary modification-substitutes HJTC number seven for number eight, to suppress the alarms. The Safety Evaluation concluded there was no unreviewed
, . safety question or a_ change in the Technical Specifications.
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10CFR50.59 Annual Esport _nagn_10
- "n-Pary Modification' 1-90-146
,nis temporary modification was to suppress nuisance alarms on Channel A of the Unit 1 Reactor Vessel Lovel Monitoring System (RVLMS) resulting from an inoperable heated junction thermocouplo (HJTC) on consor number five. The Control Room troublo and low level alarm remain in alarm. The temporary modification substitutos HJTC number six for number.five, to suppress the alarms. The Safety Evaluation concludod there was no unroviewod l safety question or a change in the Technical Specifications.
Tomoorary Modification 1-90-175 i Instrument channols 1-PT-103 and 1-PT-103-1 monitor pressurizer pressura and provido numerous automatic signals to various plant components. Unit 1 has exporlonced transient noiso on those instrumonts. Transient noiso may cauno 1-SI-651-MOV and 1-SI-652-MOV to automatically close and isolato shutdown cooling.
This temporary modification was used to temporarily defont this !
automatic' actuation signal. A dedicated control room operator was used to duplicato the function normally performed by the automatic signal. This 50.59 Safety Evaluation documents the o fact that Calvert Cliffs was temporarily in a plant configuration that did not coincide with the FSAR. The Safety Evaluation concluded-there was no unroviewed safety question or a chango in the Technical Specifications.
Temporary ModifigAtion 2-90-6l d This Safety Evaluation allows the gonoric installation of temporary-blanks or blind flangos in the Salt Water system.
Those blanks or blind flangos will allow valvos and/or spool pieces in the Salt Water System to be removed for various maintenanco activitios and still allow portior.s of the system to remain operable. Installation of blanks or blind flanges will be used to return'the Salt Water System to a condition that satisfies the operability requirements of the Technical Specifications while the affected value or spool ploco is removed
-from the system. The Safety Evaluation concluded there was no unroviewed safety question or a change in the Technical
-Specifications.=
QCR 90-472 This activity is a drawing revision to an FSAR piping and instrument diagram (PSID). M-36 shoote 1 3, and 4, will be t modiflod to show additional drain and drain valvo details for the Main Steam /Roboat systems. The drain line details being added to
-the P&ID have always existed in the plant, but woro just never explicitly shown on-the FSAR P&ID. The Satoty Evaluation
. concluded there was no unroviewed safety question or change in the Technical-Specifications.
10CPRhatE9-Annual ReDort Pace 17 l
Setpoint chance I-90-41 This Safety Evaluation was written to incorporate setpoint change I-90-41. 12 B Reactor Coolant Pump (RCP) was experiencing vibration levels of 21 mils unfiltered, causing a hanging alatu in the control Room. Increasing the current alarm setpoints from 18 mils alert, 20 mils danger to 24 mils alert, 26 mils danger, cleared the alarm and allowed operations to detect any I degradation in all four RCPs. The Safety Evaluation concluded !
there was no unreviewed safety question or chango in the '
l Technical Specifications.
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Miscellaneous Activity 50.59 #90-1-045-032-RO NCR 8584 was issued to document the concern that High' Pressure Feedwater Heaters 16 A/B have inadequate tube side over pressure i protection per ASME Section VIII. The Feed System relies on the !
Main-Feed Pump High Discharge Pressure Trip to prevent 16 A/B from an overpressure condition. FSAR Section 10.2.3 states that the heaters are designed to ASME Section VIII. However, ASME Section VIII - UG-125 does not allow crediting the main food pump trip. Therefore, High Pressure Feedwater Heaters 16 A/B are not in compliance with ASME Section VIII. Thus, operation of 16 A/B is prohibited prior to reconciling this discrepancy with both the State of Maryland and the.FSAR. This safety evaluation'was performed to allow operation of the Feed System with 16 A/B Iligh Pressure Feedwater Heaters isolated and bypassed. The Safety Evaluation concluded there was no unreviewed safety question or change in the Technical Specification.
Miscellaneous Activity 50.59 490-0-027-037-RO The purpose of this evaluation is to document the allowed radioactive contamination l'imits for the Auxiliary Boiler System.
Since.the auxiliary boiler provides essential; functions, it is
- appropriate to establish allowed contamination limits, so that operation of.the system may continue following a,possible contamination event. The-limits provide--a limit of allowable contamination' levels for normally non-contaminated systems wherein continued operation-is acceptable. .The Safety Evaluation
_ concluded there was no'unreviewed safety question or change.in
-the. Technical Specification.
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10CFR50.59 Annual Report Pace 14 Miscellaneous Activity 50.59 #90-B-052-039-R1 This change involves the une of the Containment Spray (CS) pump
+ to provide core flush and'1ong term cooling after seven days following a~LOCA. The LPSI pump is limited in operation by the effect1of radiation on the Teflon 0-ring between the gland ring and pump: body. Based on radiation levels at the -10 foot elevation of the auxiliary building reaching acceptable levels in order to make valve realignments, the switch to CS pumps can-be made after seven days, but must be done at the beginning of day
' ten. The Safety Evaluation concluded there was no unreviewed safety question or change to the Technical Specification.
Miscellaneous Activity 50.59 90-2-064-056-RO Mode 2 defines reactor start-up in which criticality is a:hieved by. withdrawing the control rods. Further reactivity control is provided-by. changing RCS boron concentration. During the trancition from Mode 2-to 100 percent power, RCS boron concentration ~is reduced substantially. This is accomplished by a dilution process that requires a significant react or coolant change-out. This dilution process is completed in npproximately 4 8. _ hours . Maintaining RCS Lithium Concentration between 1.0 -2.0 ppm during this dilution process would require excessive adds of LiOH.- This activity allows Chemistry to delay the addition of LiOH until'this dilution process is complete (provided a minimum Li concentration of .2-ppm is maintained)._ This will reduce the distraction 1to plant operators during the start-up evolution.
-This also reduces the amount-of LiOH used and the number of adds required to maintain RCS Lithium concentration. This is viewed as a1 benefit because LiOH is a caustic in the concentrated form which constitutes a-personnel safety. hazard. The proposed activity will not have a negative impact on RCS components.. The Safety Evaluation concluded there was no unreviewed safety-quesdion or'changefin the Technical Specifications.
Miscellaneous Activity 50.59 #90-B-007-008-RO As'airesult of changes to the Alternate Safe Shutdown Procedures (AOP-9), eight hour emergency light.ng is not provided-for all equipment required to be-operated fir-the procedures nor access routes there to.- This eight hour e1ergency lighting _is required by Appendix R of 10-CFR 50 bection III. J. .This_ Safety- '
Evaluation concluded there was no nnreviewed safety question due to.the'use;of compensatory measurrs until full compliance can be-Lachieved or did it require a charge to the Technical Specifications.
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10CPR50.59 Annual Report Pace 19 Miscellaneous Activity 50.59 #90-1-064-119-RO This activity allowed Reactor Coolant Pump (RCP) 12A to be operated with the seals first two stages failed, for up to fifteen minutes with Reactor Coolant System (RCS) pressure less than 330 psig and temperature less than 180 F. The reason for this activity is that RCP 12B had indication of high vibration during the initial bump and vent. It was desirable to run the ,
l pump for several minutes to see if vibration levels would i l decrease. The operating Curve requires two pumps in the same locp to be cperating. Therefore, allowing the operation of 12A RCP allowed vibration data to be obtained on RCP 12B. The Safety Evaluation concluded there was no unreviewed safety question or a ;
change in the Technical Specifications. '
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Miscellaneous Activity 50.59 #90-B-037-120-R2
- The purpose of this evaluation is to document the procer2 and I criteria used to established allowed radioactive contamination limits for ths condensate Storage Tanks (CST). The Safety l Evaluation concluded there was-no unroviewed s'iety question or a l change in the Technical Specifications.
Miscellaneous Activity 50.59 [90-B-045-129-RO This' activity evaluated the erosion of the thermal sleeve of the Steam Generator Foodwater Nozzles for the Unit 1 and Unit 2 steam 1
( generators and how this applies to the susceptibility of the nozzle to sustain a water hammer event.
This activity also L
l considers the effects of thermal fatigue and loose parts due to :
thermal sleeve erosion in the steam generators. The Safety Evaluation concluded there was no unroviewed safety question or a ,
l change in the Technical Specifications. j L hing ,'lan<m s Activity 50.59 #90-B-04)-131-RO This activity changes FSAR Table 10A-5, Instrumentation Reauired L to Place the Plant in a Safe Shutdown Condition and Maintain It
'in a Safe Shutdown Conditiont_ revising the category of some instrumentation for a critical crack in the Sample System or a letdown line break in tS Chemical and Volume Control System.
The components affected oy this change are: 1) 1-FT-212 and 1-PT-212 in the Charging Pumps discharge header; 2) 1-FT-332 and 1-Pf-
, 342 which utasure the Low Pressure Safety Injection (LPSI) flow to loops 12A and 12B; 3) 1-MOV-616, 1-MOV-626, 1-Mov-636, and 1-l L MOV-646'which are located in the High Pressure Safety Injection (HPSI) lines to loop 11A, 11B, 12A, and 128 respectively; 4) 1-FT-331 and 1-FT-341 which measure HPSI flow to loop 12A and 12B i respectively; and 5) 1-MOV-617, 1-MOV-627, 1-MOV-637, and 1-MOV-647 which are located-in the Auxiliary HPSI lines to loops 11A, l l 11B, 12A, and 12B respectively. The affected components are l being revised from Category B (qualified to be operated in a steam environment or not adversely affected by jet impingement) to Category C-(not required for a break in this system). The l
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10CFR50.59 Annual Report Pace 20
- SafetyfEvaluation con ~cluded there was no unroviewed safoty-questionior a change;in-the Technical-Specifications. 1 Miscellaneous Activity 50.59 #90-B-024-156-R2 The mainLobjectives of:this activity are to establish propor steps ~for a procedure-for restoration of Air Conditioning (A/C)
Unit No. 12 in the Main Control Room-after loss of Control Room 1
'A/C Unit No. 111due to failure of the A/C unit or failure of i Diescl' Generator-(D/G) No. 11_after a loss of offsite power to both Units,1 and to ensure that this _ activity does not constithb2 an unreviewed safety question. This scenario applios when U2.is 2 11sJdefueled;and D/G No. 21 is inoperable.- After a thorough 1 evaluation, this activity required a-chango to operating i J
procedures but did-not-constitute an unreviewed-safety question- {
or require a change to the Technical Specifications. i LTOP
- In!the LTOPcSER 8/7/78, BG&E committed to cooling the RCS using steam generators until the steam generator _.tomperaturo= ranched 220 0F. tonlyLthen would RCS cooling be switchedito the-shutdown ,
icooling system. _This.was intended to minimize the-possibility of '
starting:aireactor'coolantLpump 0 with a. secondary to primary temperature' difference of 150 F or more. "
ThisLchangeLis to procedurally allow the RCS to be cooled on the-shut-down cooling system once RCS temperaturo reaches 3000F. The maximum (steam;generatorntemperature would-then also be 3000F.
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+iAs:a'! compensatory: Peasure', all-fourLreactor.coolantLpumps (RCPs) shallohe r secured: and taggedL (breakers opened) _ prior to- cooling-the!RCS below?150 F. This ensures that no-inadvertent 1RCP start =
- will'occurswith a secondary to: primary: temperature difference-greater 1than 1150 F.: . Prior to starting an-RCP, secondary;to- 1
- primary _ delta-Tsmust be verifiediless than 150 F.
LWater= Solid operations arejnotLaffected,ERCPs are securad.andi
~ tagged Lper OP1 and OP5 7 ' no RCP " starts' are ! pormitted - durir 'ater ,
solidLoperations.
'This change. allows operations: greater; freedom _oftoperations for j theuplanti_ secondary.. . It also saves critical path ltimo'during-each= cool-downito. cold shutdown S
=The purpose:of-this evaluation was to!' document that although the proposed action =1s not in _ conformance withnour' 1978 LTOP SER, there is no.unreviewed_ safety' question-or change:to the Technical-C_..
fspecifications.
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