ML20053E342
| ML20053E342 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/12/1982 |
| From: | Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20053E326 | List: |
| References | |
| 50-322-82-04, 50-322-82-4, NUDOCS 8206070812 | |
| Download: ML20053E342 (4) | |
Text
f e
APPENDIX B NOTICE OF DEVIATION Long Island Light Company Docket No. 50-322 Shoreham Nuclear Power Station License No. CPPR-95 As a' result of an inspection conducted on February 8-26, 1982, it appears that several of your activities were not conducted in accordance with Final Safety Analysis Report (FSAR) commitments.
The following examples have been identified as deviations from the FSAR:
1.
FSAR Section 3.10.2.1.18 and Table 3.10.2.B-1 establishes approved criteria for installation of Standard Cabinets using a specified number of 5/8-inch mounting bolts.
Contrary to the above, Standard Cabinet H11*PNL-608 was installed with twenty 5/8-inch bolts instead of forty bolts and Standard Cabinets H11*PNL-635 and H11*PNL-636 were each installed with eight 5/8-inch bolts instead of twelve 5/8-inch bolts.
2.
FSAR Chapter 6.2 and Figure 6.2.5-7 describe Primary Containment Spray and specify the number of spray nozzles.
A number of drywell spray nozzles are permanently blocked by ventilation duct work, reducing the effectiveness of the containment spray system.
3.
FSAR, p.7.3-22 states that valves from other Residual Heat Removal (RHR) modes are automatically positioned so that water is correctly routed during Low Pressure Coolant Injection (LPCI) operation.
Contrary to this E11*MOV-055 and 056, one-inch RHR Heat Exchanger vents to Suppression Pool, and E11*MOV-057, RHR cooling water to Hydrogen Recombiner, are not automa'.ically positioned.
4.
FSAR Figure 7.?.1-6 and Table 7.3.2-4 describes LPCI Loop selection logic and instruments.
Contrary to this description, the logic has been deleted and is not a design feature.
5.
FSAR Table 7.3.4 shows trip set points of 2 psig for drywell pressure and 500 psig for LPCI low pressure.
Page 6.3-12 and Table 6.3.3-6 also give the LPCI low pressure set point of 500 psig.
Contrary to this the present setpoints are 1.f9 psig and 409 psig, respec-tively.
6.
FSAR Figure 7.3.1-10A&B are RHR piping and instrument drawings.
Contrary these drawings, the as-constructed plant deviates as follows:
OFFICIAL RECORD COPY 8206070812 820512 PDR ADOCK 05000322 G
Appendix B 2
Loop fill on B loop is between valves F015 and F017.
Relief valves F030A-D go to floor drains, not controlled radwaste.
Relief Valve F025 is not a thermal relief as stated in Note 12.
The line to Radwaste through valves MO-F040 and F049 is on the opposite side of valve MO-F010 as that shown.
Cooling water for RHR oumps is Reactor Building Closed Loop Cooling Water, not emergency equipment cooling water.
Drains from RHR pump suction and discharge do not tie together as shown.
7.
FSAR, p. 5.5-22 states that a relief valve on the RHR pump discharge and another on the RCIC steam supply protect the heat exchanger.
Contrary to this one relief valve is on the discharge line into the heat exchanger, with two valves intervening from the RHR pump discharge, and the steam supply is from HPCI, rather than RCIC.
8.
FSAR, p.7.3-25 states that only the air-operated check valve and check bypass valve are located in containment. Contrary to this, a manual isolation valve and manual test, vent and drain valves and connections are located in primary containment.
Your reply to this Deviation should address your plans to meet the provisions of 10 CFR 50.55(d) for bringing the original application for license up to date in its entirety at or about the time of completion of the construction of the facility.
MAY 121982 Wigizaal Signed B/ : gg Dated __
Engineering and Technical Programs OFFICIAL RECORD COPY
1 APPENDIX C OBSERVATIONS Long Island Lighting Company Docket No. 50-322 Shoreham Nuclear Power Station License No. CPPR-95 Based on the results of an inspection conducted on February 8-26, 1982, the following observations were made regarding various licensee programs. These observations are considered weaknesses in the program.
1.
No specific requirements were evident for timely incorporation of approved Engineering and Design Change Reports (E&DCR's) into drawings and specifi-cations. As an example, the two flow diagrams for the Residual Heat Removal System used for this inspection were last revised December 10, 1980. There were 34 E&DCR's outstanding against these two drawings at the time of inspection; some date back to 1978. While no violations were identified as a result of the practice, the number of E&DCR's and affected drawings and specifications lead to a concern for timely incorporation of changes as construction nears completion. The primary concern is that drawinns be completed and readily useable by plant staff for plants operations.
2.
E&DCR F-27961 established requirements for separation of class 1E and non-class 1E electrical cables in transit between raceways.
Four examples were found that did not comply with these requirements.
The Final Safety Analysis Report description of cable tray separation did not agree with recommendations of Institute of Electrical and Electronics Engineers Standard 384-1974.
The licensee is engaged in a major program to ensure adequate electrical separation throughout the plant.
The concern is that the program for ensuring electrical separation adequately address all aspects of separation, including redundancy and fire hazard considerations.
3.
Proposed Technical Specifications did not include all Residual Heat Removal pipe restraints (snubbers), did not recognize multiple snubbers and did not appropriately classify "high radiation zone" or "especially dif ficult to remove" snubbers. The proposed Technical Specifications also omitted important, plant unique, safety-related systems such as Reactor Building Closed Loop Cooling water and Low Pressure Coalant Injection Motor Generator Sets.
The concern is that submittals reflect the complete detail of the constructed plant.
4.
Carbon steel bolting used on copper nickel flanged piping, particularly the service water system, was observed to be corroded.
The condition had been identified by nonconformance reports and a corrective action plan was verbally outlined by licensee representatives.
There is a concern that the corrective action may not be thorough and may not preclude recurrence.
OFFICIAL RECORD COPY
8 s
2 You are requested to inform this office within 30 days of receipt of actions taken or planned to address these observations.
MAY 1 2 G62 origtzaal signed Bri S. Eh
[wT.T. Martin, Director,Divisionof Date v
Engineering and Technical Programs GFFICIAL RECORD COPY