ML19260D553

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Monthly Operating Rept for Jan 1980
ML19260D553
Person / Time
Site: Davis Besse Cleveland Electric icon.png
Issue date: 02/07/1980
From: Caba E
TOLEDO EDISON CO.
To:
Shared Package
ML19260D552 List:
References
NUDOCS 8002110398
Download: ML19260D553 (22)
v • d • e


Text

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AVERAGE D AILY UNIT POWER LEVEL 50-346 DOCKET NO.

Davis " ,.e Unit 1 UNIT February 7, 1980 DATE COMPLETED BY Erdal Caba 419-259-5000, Ext.

TELEPflONE 236 Jarmary, 1980 MONTil

. DAY AVERAGE DAILY POWER LEVEL

  • DAY AVERAGE DAILY POWER LEVEL (MWe-Net) (MWe-Net) 0 37 883-I 0 gg 883 2

0 19 886 3

0 20 884 4 .

0 21 883 5

0 22 880 6

45 884 7 23 g 572 24 887 820 25 884 9

887 26 885 10 882 27 885 11 887 28 884 12 887 29 880 13 881 30 001 14 883 33 881 IS 16 886 INSTRUCTIONS On this format. list the average daily unit power level in MWe-Net for each day in the reporting month. Compute to the nearest whole megawatt.

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OPERATING DATA REPORT DOCKET NO. 50-346 DATE February 7, 1980 COMPLETED BY Erdal Caba TELEPHONE A14 '50-5000, Ext.

236 OPERATING STATUS Davis-Besse Unit 1 Notes

1. Unit Name:
2. Reporting Period: January, 1980
3. Licensed Thermal Power IMb : 2772
4. Nameplate Rating (Gross 3thet: 925
5. De,ign Electrical Ratine(Net MWei: 906
6. Maximum Dependable Capacity (Gross 31We). to be deter::ine i

to be detersinctl

7. Maximum Dependa'ale Capacity (Net 31We):
8. If Changes Occur in Capacity Ratings titems Number 3 Through 7) Since Last Report. Gise Reasons:
9. Power Lese! To Which Restricted.lf Any(Net 5the):
10. Reasons For Restrictions. If Any:

This Month Yr.-to Date Cumulat;.

744 744 21,269

11. Hours in Reporting Period
12. Number Of Hours Reactor has Critical 624.1 624.1 11,588.3
13. Reactor Resene Shutdown Hours 0 0 2,875.8
14. Hours Generator On-Line 593.1 593.1 10,467.9
15. Unit Resene Shutdown Hours 0 0 1,728.2
16. Gross Thermal Energy Generated (MhH) 1,570,427 1,570,427 21,769,934 527,049 527,049 7,250,560
17. Gross Electrical Energy Generated iMWHi 18 Net Electrical Energy Generated (MWH) 496,637 496,637 6,667,215
19. Unit Senice Factor 79.7 _7_9. 7 50.6
20. Unit Asailability Factor 79.7 79.7 59.6
21. Unit Capacity Factor (Usin; MDC Net) to be determined
22. Unit Capacity ractor(Using DER Net) _73.7 73.7 37.6 20.3 . 20.3 27.2
23. Unit Forced Outage Rate
24. Shutdowns Scheduled Oser Next 6 Months (Type.Date.and Duration of Each1:

Refueling Outar,e, March, 1980 12 weeks

. 25. If Shut Down At End Of Report Period. Estimated Date of Startup:

26. Units in Test Status tPrior to Commercial Operation): Forecast Achiesed INITIAL CRITICA LITY ,

INITIAL ELECTRICITY COMMERCI \ L OPER ATION 1 GAi ) (9/77) i ,

OPERATIONAL sum!ARY JANUARY, 1980 The unit shutdown which was initiated on November 30, 1979, to investigate the motor lower bearing oil level alarm on Reactor Coolant Pump (RCP) 1-2, to fix Group 7 Rod 5 and Group 5 Rod 11 absolute position indicators ended on January 7, 1980.

1/7/80 - 1/9/80 The turbine generator was on line at 0655 hours0.00758 days <br />0.182 hours <br />0.00108 weeks <br />2.492275e-4 months <br /> on January 7, 1980. Reactor power was increased and 100% full power was achieved by 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br /> on January 9, 1980.

1/10/80 - 1/31/80 Reactor power was held at between 99% and 100% of full power for the rest of the month with the turbine generator gross load at 925 + 5 MWe.

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DATE: January, 1980 REFUELING INFORMATI0'{

Name cf facility: Davis-Besse Nuclear Power Station Unit 'l 1.

. March, 1980

2. Scheduled date f or next refueling shutdown:

June. 1980

3. Scheduled date for restart following refueling:
4. Vill refueling e r resemption of operation thereaf Ifteranswer requireisayes, technical what, specification change or other license amendment?If answer is no, has the reload in general, will these be?

and core configuratica been reviewed by your Plant Safety Review Ccomittee h to determine whether any unreviewed safety questions are associated wit the core reload (Ref.10 CFR Section 50.59)?

Yes, see att.iched

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5. Scheduled date(s) for submitting proposed licensing action and supporting December, 1979 information.

e.g., new or

6. Important licensing considerations associated with refueling, different fuel design or supplier, unreviewed design or perf ormance analysis methods, significant changes in fuel design, new operating procedures.

The spent fuel pool capacity expansion program was approved by the NRC 1, 1979.

in Amendment 19 to the operating license received August (a) in the core and (b) in the spent fuel

7. The number of fuel assemblics storage pool.

(b) . 0 (zero) 177 (a)

8. The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned, in numbe,r of fuel assemblics.

Increase size by 0 (zero)

Present 735

9. The projected date of the last refueling that can be discharged
  • to the spent ,,

fuel pool assuming the present licensed capacity.

fuel 1989 (assuming ability to unload the entire core into the spent Date pool is .maintainen and the unit goes to an lo month retueling cyclej sn?\ \ 7 l\

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REFUELING INFORMATION Continued Page 2 of 2

4. See BAW-1598 DB-1, Cycle 2 Reload Report, January,1980 The following Technical Specifications (Part A) will require revision:

2.1.1 & 2.1.2 - Reactor Core Safety Limits (and Bases) 2.2.1 - Reactor Protection System Instrumentation Setpoints (and Bases) 3.1.3.6 - Regulating Rod Insertion Limits 3.1.3.7 - Rod Program 3.1.3.9 - Axial Power Shaping Rod Insertion Limits 3.2.1 - Axial Power Imbalance (and Bases) 3.2.5 - DNB Parameters (and Beses)

The following Technical ~ Specifications (Part A) may also require revision:

3.1.2.8 & 3.1.2.9 - Borated Water Sources (and Bases) 3.2.4 - Quadrant Power Tilt (and Bases) 3.4.1 - Reactor Coolant Loops r

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 78-093 SYSTEM: Steam and Feedwater Rupture Control System (SFRCS)

COMPONENT: Controls for Main Steam Isolation Valves (MSIVs)

CHANGE, TEST, OR EXPERIMENT: On August 8, 1979 post implementation testing for FCR 78-093 was completed. This FCR changed the SFRCS trip circuitry for the MSIVs and for the Auxiliary Feedwater Isolation Valves so that the occurrence of a SFRCS half trip does not latch in and remain in the trip state after the half trip signal is removed. The occurrence of a full trip signal, however, still latches or " seals" in the full trip state. This change was made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: With the original arrangement of the system, a half trip would seal in during testing; then a spurious half trip becurring before the previous half trip generated by the testing is reset) would cause a SFRC3 full trip. Two non-The above change rectifies concurrent spurious signals would cause the same results.

this problem.

SAFETY EVALUATION: This FCR changes the controls for the main steam. isolation valves (MS100 and MS101) so that a full trip seals in, but a half trip will not seal in.

Similarly, the controls for the auxiliary feedwater isolation valves (AF599 and AF608) are changed so that a full trip seals in, but a half trip will not seal in. The changes will improve the system, and will not adversely affect the safety of the plant.

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COMPI.ETED FACILITY CllANCE REQUEST FCR No: 78-456 SYSTEM: Makeup and Purification COMPONENT: Makeup Tank CHANGE, TEST, OR EXPERIMENT: On June 30, 1979 the installation and testing of local level instrumentation on the makeup tank was completed as requested by FCR 78-456.

The level indicator, installed in the makeup pump room, is non-nuclear safety related, llowever, this FCR is nuclear safety related by virtue ot the fact that post inspec-tion construction authorization (PICA) review was required to ensure that the instal-lation did not create an environment adverse to safety. Procedural changes to reflect this change have been incorporated in Revision 8 to EP 1202.33, " Emergency Operation of the NSSS".

REASON FOR THE FCR: In the event of a control room evacuation, there is no means of determining makeup tank level. With no local level indication, it is possible that the makeup pumps would trip on low makeup tank level.

SAFETY EVALUATION: Portions of this FCR require PICAS. Implementation of those portions of the work packages are nuclear safety related in that they are non-Q-list items which could have an indirect effect on the function of Q-list structures, systems, or components. Unplementation in accordance with the PICAS will preclude these items from affecting the safety function of the Q-list structures, systems, or components. No unreviewed safety question is involved.

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COMPLETED FACILITY CHANGE REQUEST FCh NO: 79-092 SYSTEM: Fire Protection COMPONENT: Additional fire extinguishers CHANGE, TEST, OR EXPERIMENT: On December 21, 1979, work was completed under FCR 79-092 to install 23 additional fire extinguishers of various types in the Auxiliary Building and the intake structure. This FCR is nuclear safety related by virtue of the fact that Post Inspection Construction Authorization (PICA) review was required to ensure that the installation of the additional fire extinguishers would not create an environment potentially adverse to safety.

REASON FOR THE FCR: The installation of these fire extinguishers was committed to in the Davis-Besse Unit 1 Fire Hazard Analysis Report.

SAFETY EVALUATION: This change is classified as nuclear safety related because PICAS are required to ensure that no new adverse environments are created.

Installation in accordance with PICAS will ensure this. This is therefore not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-347 SYSTEM: Component Cooling Water (CCW) System COMPONENT : Decay Heat Removal Coolers CCW Outlet Valves CC1467 and CC1469 CHANGE, TEST, OR EvPERIMENT: On December 10, 1979, implementation of Facility Change Request 79-347 was completed. This change modified the actuator linkages on valve CC1467 and valve CC1469 by replacing the cap screw holding slide clamp and the valve disc arm together with a screw penetrating the entire thickness of the disc arm. This provides a more secure and positive attachment of the actuator linkage arms to the disc arms of the valves. This change is identical to that implemented under FCR 79-151 on Service Water Valves SW1424, SW1429 and SW1434, which have the same type of actuator.

REASON FOR THE FCR: The retaining screw for the valve linkage was being loosened by vibration causing slippage and misalignment of the valve operator linkage, see Licensee Event Reports NP-33-79-145, NP-33-79-ll4, and NP-33-79-94.

SAFETY EVALUATION: The proposed solution to the actuator slippage problem des-cribed in FCR 79-347 will in no way impair the ability of valves CC1467 and CC1469 to perform their intended safety function. The proposed solution should increase the reliability of the valves by preventing linkage arm slippage and ensuring proper alignment between the actuator and the valve. No unreviewed safety question is created by this FCR. .

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COMPLETED FACILITY CHANGE REOUEST FCR No: 79-376 SYSTEM: Reactor Protection System COMPONENT: Nuclear Instrument (NI) 1 CHANGE, TEST, OR EXPERIMENT: On November 11, 1979 implementation of FCR 79-376 was completed. Under this FCR, the three cable schemes associated with the source range indication of NI-l (2LRPSA03X, 2LRPSA03Y, and 2LRPSA03Z) were relocated from conduit 39034A to the previously spare conduit 39035A. New cable was used for these schemes.

The three other cable schemes located in conduit 39034A were not affected.

REASON FOR THE FCh: The station had been experiencing repetitive failures of source range detector NI-l due to an unknown cause (see Licensee Event Reports NP-33-79-107, NP-33-79-87, NP-33-78-110, NP-33-79-100, NP-33-77-88, and NP-33-77-79). Trouble-shooting performed by Westinghouse field engineers using a time domain reflectometer indicated that the problem probably lies in the NI-l source range cabling. For this reason, the NI-l source range cabling was replaced and isolated from the power range cabling of NI-5 as detailed above. It is expected that this change will improve NI-l reliability.

SAFETY cVALUATION: Conduit 39034A contains presently the following cables: 2LRPSA01X, 2LRPSA01Y, 2LRPSA0lZ, 2LRRSA03X, 2LRPSA03Y, 2LRPSA03Z. The subject FCR desires to relocate the latter three cables schemes in existing spare conduit 39035A.

Both conduits 39034a and 39035A originate, terminate, and run together. They also are both 2", PVC coated, rigid steel construction. It is, therefore, acceptable to relocate the three cables schemes. The health and safety of the public, platit personnel and equipment are not affected by this change. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR No: 79-379 SYSTEM: Service Water COMPONE'.T : Seismic langer 41-HBC-44-H5 CilANGE, TEST, OR EXPERIMENT: Impicmentation of FCR 79-379 was completed on November 19, 1979. This FCR added cover bars to the channti members of seismic support 41-11BC-44-115 in order to reduce the anticipated stcess which would occur in the hanger when under the design basis loading. The hanger is on the service water supply piping to Emergency Core Cooling System Room Cooler 1-3. The change reinforced the hanger to increase the safety factor in order to meet the NRC criteria for pipe support operability. This change was made under the direction of the unit architect /

engineer, Bechtel Company.

REASON FOR THE FCR: It was found during a review for IE Bulletin 79-07 that the design of this and several other scismic hangers were not as conservative as assumed. Hanger 41-ilBC-44-H5 did not meet the NRC design criteria for pipe support operability; the safety factor was less than 2. See Licensee Event Report NP-32-79-13 for further details.

SAFETY EVALUATION: The addition cf the welded cover bar to the existing hanger will create a box structure. The hanger strength will be increased and the stresses will be lowered to ac:eptable limit 9, including a safety factor. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR No: 79-380 SYSTEM: Emergency Core Cooling System (ECCS) 1 COMPONENT: Seismic hanger 33A-HCB-2-H44 CHANGE, TEST, OR EXPERIMENT: On November 19, 1979, implementation of FCR 79-380 was completed. This FCR added a 3" x 3" x 3/8" angle iron diagonal brace to scis-mic hanger 33A-HCB-2-H44 in order to reduce the potential stress in the hanger in order to increase the saf2ty factor as required to meet the NRC criteria for pipe support operability. This hanger is on the 14" line Thiswhich is the common borated modification was made under water supply line for all the ECCS 1-1 equipment.

the guidance of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR:

A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other seismic hangers was not as conservative Hanger as required by ti.e design criteria and assumptions used by the hanger vendor. The 33A-HCB-2-H44 did not meet the design criteria for pipe support operability.

above modification corrects this deficiency on this hanger. See Licensee Event Report NP-32-79-13 for further details.

SAFETY EVALUATION: The addition of the welded angle member at 45 to the existing This horizontal and vertical menbers will reduce the stress to acceptable levels.

is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REOUESTS FCR NO- 79-381 b;3 TEM: Containment Spray COMPONENT: Scismic hangers 34-CCB-5-H17 and 34-HCC-38-H19 CHANCE, TEST, OR EXPERIMENT: On November 19, 1979 implementation of FCR 79-381 was c omple t ed. Under this FCR, 2 " pipe and 3/4" x 7" gusset plates were added to the above seismic hangers to stiff en the hangers and thus decrease the anticipated hanger deflection under the design basis loading. Hanger 34-GCB-5-H17 is located on the discharge line of containment spray pump 1-1 and hanger 34-HCC-38-H19 is lo-cated on the recirculation test line of containment spray pump 1-1. These changes were mad; under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: During a review conducted for a response to IE Bulletin 79-02, it was discovered that the design of this and several other hangers was not as con-servative as required by the canuf acturer's design criteria. It was found that on these two supports the calculated deficction under the design basis loading would have exceeded the vendor's design criteria of 0.0625". See Licensee Event Report NP-32-79-13 for further details.

SAFETY EVALUATION: The addition of pipe welded to these supports will stiffen the hangers and decrease the deficction. The change will not have any adverse effect on the safety function of the ..ontainment spray system. This is not an unreviewed safety question.

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COMPLETED FACILITY CllANGE REQUESTS FCR No: 79-387 SYSTEM: Main Steam COMPONENT : Seismic hanger 3A-EBD-19-Il78 CllANGE, TEST, OR EXPERIMENT: On November 19, 1979, implementation of FCR 79-387 was completed. This FCR added stiffeners, new braces, and base plates to seismic hanger 3A-EBD-19-Il78 which is on the main steam line to Auxiliary Feedpump Turbine 1-2.

This modification was done under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR TIIE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other hangers was not as conservative as required by the manufacturer's design criteria. See Licensee Event Report NP-32-79-13 for further details. The modifications made to this hanger reduced the slenderness ratio to under 200 as required.

SAFETY EVALUATION: The addition of the stiffener plates, new braces, and base plates will reduce the slenderness ratio to an acceptable level. The changes will not have an adverse effect on the main steam system. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-388 SYSTEM: Auxiliary Feedwater COMPONENT: Seismic hanger 6C-EBD-14-H43 CHANCE, TEST, OR EXPERIMENT: On November 19, 1979, a modification to seismic hanger 6C-EBD-17-H43, located on the discharge piping of Auxiliary Feedwater Pump 1-2 was completed as requested by FCR 79-388. The modification was the replacement of the diagonal brace or kicker with 4" x 4" x 3/8" structural steel tubing in order to reduce the slenderness ratio of the hanger to a level in compliance with the manu-facturer's design criteria of a slenderness ratio less than 200. This change was made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other hangers was not as conservative as required by the manufacturer's design criteria. See Licensee Event Report NP-32-79-13 for further details. The modifications made to this hanger reduced the slenderness ratio to under 200 as required.

SAFETY EVALUATION: The addition of the angle brace will reduce the slenderness ratio to an acceptable icycl. The changes will not have an adverse effect on the auxiliary feedwater system. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE P90UESTS FCR No: 79-389 SYSTEM: Makeup and Purification COMPONENT: Seismic hanger 31-CCB-21-H22 CHANCE, TEST, OR EXPERIMENT: On November 19, 1979 a modification to seismic hanger 31-CCB-21-H22 was completed as requested by FCR 79-389. The modification consisted of the addition of a 1/2" r. 5" cover bar to increase the saf ety factor of the hanger in order to meet the NRC criteria for pipe support operability. This modification was made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR T!!E FCR: During a review for IE Bulletin 79-02, it was discovered that the design of this and several other seismic hangers was not as conservative as re-quired by the design criteria and assumptions used by the hanger vendor, ITT Grinnell.

Hanger 31-CCB-21-H22 and three others did not meet the NRC design criteria for pipe suppert operability. See Licensee Event Report NP-32-79-13 for further details.

SAFETY EVALUATION: The addition of stiffener plates will reduce the stress in the hanger to acceptable levels. It will not have an adverse ef fect on the makeup and purification (letdown portion) system. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUEST FCR NO: 79-390 SYSTEM: High Pressure Injection COMPONENT: Seismic hanger 33A-GCB-4-H5 CRANGE, TEST, OR E}?ERIMENT: On November 19, 1979 stiffener plates were added to the channel brace of seismic pipe hanger 33A-GCB-4-H5 as requested by FCR 79-390.

This modification reduces the slenderness ratio of the hanger to a value less than 200 to comply with the vendor's design criteria. This modification was made under the direction of the architect / engineer, Bechtel Company.

REASON FOR THE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other hangers was not as conservative as re-quired by the manuf acturer's design criteria. See Licensee Event Report NP-32-79-13 for further details. The modifications made to this hanger reduced the slenderness ratio to under 200 as required.

SAFETY EVALUATION: The addition of stiffener plates to the channel brace will reduce the slenderness ratio to an acceptable level. This change will not have an adverse effect on the high pressure injection system. This is not an unreviewed safety question.

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COMPLETED FACILITY CRANGE REQUESTS FCR NO: 79-391 SYSTEM: Containment Spray COMPONENT: Seismic hanger 34-CCB-5-H2 CHANGE, TEST, OR EXPERIMENT: On November 19, 1979 a 3/8" x 41" 3 stiffener plate was added to the brace of seismic pipe hanger 34-GCB-5-H2, which is located on the dis-charge piping of containment spray pump 1-2, tras completed as requested by FCR 79-391. This modification reduces the slenderness ratio of the support to a value less than 200 in compliance with the hanger vendor's design criteria. This modifi-cation was made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other hangers was not as conservative as required by the manufacturer's design criteria. See Licensee Event Report NP-32-79-13 for further details. The modifications made to this hanger reduced the slenderness ratio to under 200 as required.

SAFETY EVALUATION: The addition of the :tiffener plates to the angle brace I-beam will reduce the slenderness ratio to an acceptable level. This change will not have an adverse effect on the containment spray system. This is not an unreviewed safety question.

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COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-392 SYSTEM: Component Cooling Water (CCW)

COMPONENT: Seismic hanger 36-HBC-39-H8 CHANCE, TEST, OR EXPERIMENT: On November 15, 1979, a modification to seismic hanger 36-HBC-39-H8 was completed as requested by FCR 79-392. The hanger is located on the CCW supply line to the letdown coolers. The modification consisted of the addition of stiffener plates to the diagonal brace (kicker) in order to reduce the slender-ness ratio of the hanger to a value less than 200 to comply with the vendor's design criteria. This change was made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other hangers was not as conservative as required by the manufacturer's design criteria. See Licensee Event Report NP-32-79-13 for further details. The modifications made to this hanger reduced the slenderness ratio to under 200 as required.

SAFETY EVALUATION: The addition of stiffener plates to the vertical support and angle brace 1-beams will reduce the slenderness ratio to an acceptable level. This change will not have an adverse effect on the CCW system. This is not an unreviewed safety question.

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l COMPLETED FACILITY CHANGE REQUESTS FCR NO: 79-393 SYSTEM: Service Water COMPONEITr: Seismic hanger 41-HBC-36-H26 CHANGE , TEST, OR EXP ERIMENT: Implementation of FCR 79-393 was completed on November 19, 1979. This FCR added web stiffeners and a wide flange beam to seismic support 41-HBC-36-H26 which is on the service water outlet piping on component cooling water heat exchanger 1-3. This reinforcement of the support increases that safety factor to comply with the NPC criteria for pipe support operability. This modification was made under the guidance of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: A review conducted for a response to IE Bulletin 79-02 found that the design of this and several other seismic hangers was not as conservative as required by the vendor's design criteria. Hanger 41-HBC-36-H26 did not meet the NRC criteria for pipe support operability as the factor of safety was less than 2. The above modification corrects this deficiency for this hanger. See Licensee Event Report NP-32-79-13 for further details.

SAFETY EVALUATION: The addition of web stiffeners and wide flange beam to the hanger will reduce the stress to acceptable levels. This change will have no adverse effect on the service water system. This is not an unreviewed safety question.

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4 COMPLETED FACILITY CHANGE REQUEST FCR No: 79-445 SYSTEM: High Pressure Injection (HPI)

COMPONENT: Seismic hanger 33C-CCB-2-H46 CHANGE, TEST, OR EXPERIMENT: On December 23, 1979, modifications were completed on seismic hanger 33C-CCB-2-H46 as requested by FCR 79-445 including Supplement 1.

This hanger is located on HPI line 1-1. The modifications involved the replacement of the angle iron members of the support with structural tubing thus reducing the slenderness ratio to less than 200 as required by the vendor's design criteria.

These changes were made under the direction of the unit architect / engineer, Bechtel Company.

REASON FOR THE FCR: During the conduct of a review being made for IE Bulletin 79-02, it was discovered that the design of hanger 33C-CCB-2-H46 exceeded the slen-derness design criteria of 200 used by the hanger and piping vendor (see Licensee Event Report NP-33-79-154 for further details). The above modification corrects the design of this hanger.

l SAFETY EVALUATION: The modifications to this hanger will assure that the design criteria slenderness ratio of 200 is satisfied. This will assure hanger performance in accordance with design criteria. There will be no adverse ef f ects on the HPI system performance. Therefore, this is not an unreviewed safety question.

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