ML20197B693

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10CFR50.59 Annual Rept of Changes,Tests & Experiments for 841001-860322
ML20197B693
Person / Time
Site: Diablo Canyon  Pacific Gas & Electric icon.png
Issue date: 03/22/1986
From:
PACIFIC GAS & ELECTRIC CO.
To:
Shared Package
ML20197B679 List:
References
NUDOCS 8610300378
Download: ML20197B693 (9)
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PGandE Letter No.: DCL-86-300 ENCLOSURE DIABLO CANYON UNITS 1 AND 2 DOCKET NOS. 50-275 AND 50-323 10 CFR 50.59 ANNUAL REPORT OF CHANGES, TESTS, AND EXPERIMENTS OC(0BER 1, 1984 - MARCH 22, 1986 Pacific Gas and Electric Company l

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8610300378 861017 PDR ADOCK 05000275 R PDR 1103S/0047K

ATTACHMENT 1 5

SUMMARY

OF FACILITY CHANGES FOR ANNUAL REPORT 2

October 1, 1984 - March 22, 1986 DCPP Unit Change HQ1 Subiect Acolicability Identification

1. Post-Accident Neutron Flux Monitors 1, 2 DC1-EE-13979 (RG 1.97) DC1-EJ-15696 DC2-EJ-14979
2. Electrical Penetrations (RG 1.97) 1 DCI-EC-19688
3. Accumulator Wide Range Level Indicators 1, 2 DCl-EJ-17998
(RG 1.97) DC2-EJ-18998 l
4. Containment Atmosphere and Sumps 1, 2 DCl-EJ-27311 Temperature Monitors (RG 1.97) DCP-J-27357 DC2-EJ-28311 DC2-EJ-18998
5. Containment Sump Level Recorder 1 DC1-EJ-27702
6. Containment Spray Flow Indication 1, 2 DCl-EE-27355 (RG 1.97) DC2-EJ-18998
7. Liquid Holdup Tanks Level Indicators 1, 2 DC1-EJ-27356 (RG 1.97) DC2-EJ-18998
8. Extended Temperature Monitor Range for 1, 2 DCP-J-27361 Pressurizer Relief Tank (RG 1.97) DC2-EJ-18998
9. Qualified Limit Switches on Containment 1 DCP-J-27359 Isolation Valves (RG 1.97) DC1-EJ-29736 DCl-EJ-27359
10. Steam Generator Blowdown Bypass 2 DC2-EM-24151
11. D/G 1-3 Control Switches 1, 2 DC0-EE-25999
12. Differential Pressure Transmitters Inside 2 DC2-EJ-12325 Containment
13. Clarify Containment Isolation Signals 1, 2 DCO-EE-15562
14. Redundant CVCS Letdown Filter 2 DCP-M-28132 6

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1103S/0047K 1-1

FACILITY CHANGES FOR ANNUAL REPORT October 1, 1984 - March 22, 1986

1. Post-Accident Neutron Flux Monitors (RG 1.97)

DC1-EE-13979, DC2-EJ-14979 and DC1-EJ-15696 (Units 1 and 2)-

In accordance with the requirements of Reg. Guide 1.97, these changes provide permanent, wide range, qualified neutron flux monitors for operator information only. These changes also aid plant compliance with 10 CFR 50, Appendix R. Neutron flux indication is provided in the control room and remote shutdown panel.

Safety Evaluation Summary These changes provide an additional source of operator information regarding the criticality of the core. These changes do not affect any control or automatic safety functions.

2. Electrical Penetrations (RG 1.97)

DCl-EC-19688 (Unit 1)

Two new electrical penetration assemblies are added to existing spare penetration locations to support the addition of new neutron flux monitors required per Reg. Guide 1.97.

Safety Evaluation Summary This change is required to conform to Reg. Guide 1.97. The new assemblies are designed, constructed, installed, and tested to previously approved codes, standards and procedures.

3. Accumulator Hide Ranae Level Indicators (RG 1.97)

DC1-EJ-17998 and DC2-EJ-18998 (Units 1 and 2)

This change adds wide range water level indicators for the accumulator tanks in accordance with Reg. Guide 1.97 post-accident monitoring '

requirements.

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Safety Evaluation Summary This change provides additional information to operators regarding the l status of the accumulators. No new automatic equipment actions or i failure modes are involved.

Il03S/0047K 1-2

4. Containment Atmosohere and Sumos Temoerature Monitors (RG 1.97)

DCl-EJ-27311, DC2-EJ-28311, DCP-J-27357, and DC2-EJ-18998 (Units 1 and 2)

These changes add temperature monitors in the containment atmosphere, reactor cavity sump, and containment recirculation sump in accordance with the post-accident monitoring requirements of Reg. Guide-l.97.

Safety Evaluation Summary The new monitors provide control room indication of containment atmosphere and sump temperature for operator information only. This change does not affect any control functions nor is any new failure mode created.

5. Containment Sumo Level Recorder DCl-EJ-27702 (Unit 1)

This change replaces three level indicators with one level recorder in the control room for the reactor cavity sump and containment recirculation sump. The purpose of this change is to conserve space on the Post Accident Monitoring panel.

Safety Evaluation Summary The addition of a three channel level recorder offers net improvement in operator information presentation due to trending enhancement.

6. Containment Soray Flow Indication (RG 1.97)

DCl-EE-27355 and DC2-EJ-18998 (Units 1 and 2)

This change adds flow monitors to the containment spray pump discharges and flow indication in the control room to provide operators with information regarding containment spray operation. This change is required per Reg. Guide 1.97.

Safety Evaluation Summary The flow monitors utilize existing flow orifices in the pump discharge lines. This change provides operator information only and does not affect any control functions.

7. Liauid Holduo Tanks Level Indicators (RG 1.97)

DCl-EJ-27356 and DC2-EJ-18998 (Units 1 and 2)

This change adds wide range level instrumentation to the liquid holdup tanks to meet the requirements of Reg. Guide 1.97.

Safety Evaluation Summary This new instrumentation provides operator information only and uses existing tank taps.

Il03S/0047K 1-3

8 Extended Temoerature Monitor Ranae for Pressurizer Relief Tank (RG 1.97)

DCP-J-27361 and DC2-EJ-18998 (Units 1 and 2)

These changes expand the pressurizer relief tank (PRT) temperature indicator range to 50-350*F to partially comply with Reg. Guide 1.97.

Safety Evaluation Summarv -

Reg. Guide 1.97 requests an indicator range of 50 to 750*F for the PRT.

However, accuracy at this extended range is lessened so that small changes in temperature, indicative of a leaking relief valve, could not be detected by the operators. Consequently, the range was expanded to 50-350*F to provide the desired accuracy.

9. Qualified Limit Switches on Containment Isolation Valves (RG 1.97)

DCl-EJ-29736, DCP-J-27359, and DCl-EJ-27359 (Unit 1)

These changes add qualified limit switch (es) on inside containment isolation valves to ensure operator knowledge of post-accident valve position.

Safety Eypluation Summaty Upgraded components are required in accordance with Reg. Guide 1.97, Ittm

15. These changes have been implemented in accordance with existing approved codes and standards.
10. Steam Generator Blowdown Bvoass DC2-EM-24151 (Unit 2)

This chtnge allows manual diversion of steam generator blowdown to the main condenser from the blowdown demineralizer system.

Safety Evaluation Summary Manual diversion is placed under administrative control so that no diversion will be allowed if radioactivity in blowdown water exceeds allowed levels.

11. D/G 1-3 Control Switches )

DCO-EE-25999 (Units 1 and 2) 1 This change revises auto / manual switch logics so that one unit can retain ;

auto-start operation while the other unit has its control switch in i manual. The diesel generator will load to the unit in " auto" upon automatic demand.

S3fety Evaluation Summary l

This change allows more flexibility in operations and maintenance by allowing a vital bus outage in one unit (e.g., in Modes 5 or 6) while the swing diesel can still be operable for the other unit.

1103S/0047K 1-4

12. Differential Pressure Transmitters Inside Containment DC2-EJ-12325 (Unit 2)

This change replaces certain Barton differential pressure transmitters used in safety-related applications inside containment with qualified transmitters supplied by Rosemount.

Safety Evaluation Summary New components meet or exceed all requirements of original design. Minor changes in mounting and/or connection details were revised to support new components.

13. Clarifv Main Steam and Feedwater Isolation Sianals DCO-EE-15562 (Units 1 and 2)

This change clarifies actual containment isolation signals to close main steam valves and feedwater valves.

Safety Evaluation Summary Isolation signal logic and hardware unchanged from approved configuration; isolation logic drawings restated to represent clarified hardware configurations.

14. Redundant CVCS Letdown Filter DCP-M-28132 (Unit 2)

This change adds a parallel, redundant filter for reactor coolant letdown to the volume control tank to allow catching demineralizer debris even with a filter out of service.

Safetv Evaluation Summary To maintain piping integrity, the new piping and supports are designed and constructed to equivalent codes as the original filter. ALARA considerations are incorporated into the design. No net increase in radwaste system inputs is expected.

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1103S/0047K 1-5

ATTACHMENT 2 PROCEDURE CHANGES FOR ANNUAL REPORT October 1, 1984 - March 22, 1986 For a procedure to be " described" in the FSAR:

1. The FSAR verbiage must contain a commitment to perform a manual action or establish an administrative control or program.
2. The FSAR verbiage must contain a detailed description of the methodology used to perform these activities with mention of:

a) equipment used (as appropriate),

b) limits on the activity, and/or c) order of steps.

3. An FSAR system description that includes an explanation of how plant personnel can interact with the system does not necessarily constitute a requirement for a procedure.
4. The activities described in the FSAR must be significant such that they could affect the safety of the plant.

Procedure changes pertain to procedural commitments and to procedures discussed in the initial tests and operations sections of the FSAR (Section 14).

A. Procedural Commitments Many procedural commitments are described in the FSAR and are incorporated into DCPP procedures such as:

Nuclear Plant Administrative Procedures (NPAP)

  • Operating Procedures (0P)
  • Emergency Procedures (EP)

Mechanical Maintenance Procedures (MMP)

Surveillance Test Procedures (STP)

Radiation Control Procedures (RCP)

  • Chemistry and Radiation Procedures (CRP)
  • Quality Control Procedures (QCP)
  • Inservice Inspection (ISI) Procedures
  • Inservice Inspection (ISI) Procedures
  • Annunciator Response (AR) Guidelines The PSRC has reviewed all administrative and technical procedure revisions and concluded that they do not present unreviewed safety questions or changes to the DCPP Technical Specifications. Rone of the revised procedures resulted in a deviation frcm the steps listed in the FSAR or resulted in a system operation that deviated from the way that system was described in the FSAR. Therefore, no procedure changes are reportable under 10 CFR 50.59.

B. Initial Tests and Ooerations ,

Changes to the DCPP Unit 2 initial test program described in Section 14 of the FSAR have been made in accordance with the provisions of 10 CFR 50.59. Pursuant to license condition 2.C.(3) of DPR-82, these changes have been previously reported to the NRC in accordance with 10 CFR 50.59(b). None of these changes involve an unreviewed safety question or a change to the DCPP Technical Specifications, as determined by the PSRC. These changes are described below, including the PGandE letter that documented the change and a safety evaluation summary.

Pursuant to license condition 2.C.(3) of DPR-80, changes to the DCPP Unit 1 initial test program described in FSAR Section 14 have received prior NRC approval and are not considered 50.59 changes.

1. Deletion of Certain Unit 2 Radiation Survev and Shieldina Effectiveness Tests (PGandE Letter DCL-85-240, dated July 16, 1985)

This change deleted the radiation survey tests at the 30, 75, and 90 percent power levels from the Unit 2 power ascension program.

Safety Evaluation Summary Deletion of these surveys is not detrimental to determining shielding accuracy, and it results in less occupational radiation exposure to personnel performing the surveys.

2. Deletion of Unit 2 Static Rod Droo and RCCA Below Bank Position Measurement Test (PGandE Letter DCL-85-354, dated November 25, 1985)

This change deleted the Static Rod Drop and RCCA Below Bank Position Measurements Test from the Unit 2 power ascension program.

Safety Evaluation Summary Deletion of this test does not affect the DCPP FSAR rod drop accident analysis.

Il03S/0047K 2-2

3. Deletion of Unit 2100 Percent Net Load Trio Test (PGandE Letter DCL-85-098, dated April 11, 1986)

This change removes the requirement to successfully complete the performance of a 100 percent net load rejection test during the Unit 2 power ascension program.

Safety Evaluation Summary Deletion of this portion of the testing does not involve a significant increase in the probability or consequences of an accident previously evaluated since credit for sustaining a loss of external load without reactor trip was not assumed in the accident analysis. Deletion of this testing does not create the possibility of a new or different kind of accident from any accident previously evaluated since it does not involve a physical alteration of the plant or affect normal plant operation. Deletion of the testing does not involve a reduction in the margin of safety as defined in the bases for any unit Technical Specification or in the Unit 2 accident analysis.

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1103S/0047K 2-3

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