ML16341C579
| ML16341C579 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/31/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0675, NUREG-0675-S27, NUREG-675, NUREG-675-S27, TAC-51994, NUDOCS 8408160250 | |
| Download: ML16341C579 (34) | |
Text
NUREG-0675 Supplement No. 27 Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2 Docket Nos. 50-275 and 50-323 Pacific Gas and Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation July 1984
1
ABSTRACT Supplement 27 to the Safety Evaluation Company's application for a license to Plant, Unit 1 (Docket No. 50-275),
has Reactor Regulation of the U.S.
Nuclear addresses the revisions to the license cations as they relate to Amendment 10 ating License DPR-76.
Report for Pacific Gas and Electric operate Diablo Canyon Nuclear Power been prepared by the Office of Nuclear Regulatory Commission.
This supplement conditions and to the Technical Specifi-to Diablo Canyon, Unit 1 Facility Oper-Diablo Canyon SSER 27
I'
TABLE OF CONTENTS ABSTRACT..
~Pa e
INTRODUCTION REVISED LICENSE CONDITIONS l.
2.
3.
4.
5; 6.
7.
8.
9.
10.
11.
Maximum Power Level Fire Protection System (Appendix R to 10 CFR Emergency
Response
Capabilities Masonry Walls..
Seismic Design Bases Revalidation Program...
Overhead Heavy-Load Handling Systems Reporting of Violations.....................
Emergency Preparedness....,...
Jet Impingement......,,,...,............"..
Piping and Piping Supports License Expiration Date..
50)
~
~
~
~
~
~
~
~
~
~
~
1 1
1 1
1 1
1 1
2 2
2 REVISED TECHNICAL SPECIFICATIONS l.
2.
3.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
23.
24.
Plans and Accident Monitoring Instrumentation Combustible Gas Control Diesel Generator Testing Containment Ventilation System..........,......
Containment Ventilation System Reactor Vessel Head Vents Testing of HEPA Filters and Charcoal Filters Moderator Temperature Coefficient (MTC)...
Pressurizer Code Safety Valves Reactor Trip Surveillance Fan Coolers Snubbers Flood Protection (Breakwater)..
Fire Protect>on Limits on Working Hours Frequency of Auditing Emergency and Security Implementing Procedures Administrative Controls Natural Circulation and Boron Dilu'tion Test..
Overtemperature bT Reactor Coolant System MODE 3..
D.C.
Sources Containment Free Volume NRC Organizational Changes High Radiation Area 2
"2 2
2 2
3 3
3 3
3 3
3 3
3 4
4 4
4 4
Diablo Canyon SSER 27
TABLE OF CONTENTS (Continued)
IV.
DISCUSSION OF REVISED LICENSE CONDITIONS AND RESOLUTION OF ISSUES
~Pa e
l.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
Maximum Power Level Fire Protection System (Appendix R to 10 CFR 50)
Emergency
Response
Capabilities Masonry Walls....,
Seismic Design Bases Revalidation Program....
Overhead Heavy-Load Handling Systems....
Reporting of Violations Management of Operations (Section I.B. 1)
Jet Impingement.......
Piping and Piping Supports License Expiration Date................
Compliance with Regulatory Guide 1.97...
Low Flow Alarm......
Low Temperature Overpressure Protection.........
4 5
7 7
8 10llll 12 12 12 12 13 13 V.
DISCUSSION OF REVISED TECHNICAL SPECIFICATIONS.......
l.
2.
3.
4.
5.
6.
7.
8.
9.
10.
11.
12.
13.
14.
15.
16.
17.
18.
19.
20.
21.
22.
23.
24.
Accident Monitoring Instrumentation........
Combustible Gas Control Diesel Generator Testing................
Containment Ventilation System..
Containment Ventilation System Reactor Vessel Head Vents ;.................
Testing of HEPA Filters and Charcoal Filters Moderator Temperature Coefficient (MTC)
Pressurizer Code Safety Valves......
Reactor Trip Surveillance........
Fan Coolers......................
Snubbers.......
Flood Protection (Breakwater)..
Fire Protection....
Limits on Working Hours.....
Frequency of Auditing Emergency and Security Implementing Procedures Administrative Controls................
Natural Circulation and Boron Dilution Test Overtemperature hT..........................
Reactor Coolant System MODE 3...
D.C.
Sources................................
Containment Free Volume NRC Organizational Changes High Radiation Area...,............'.......
Plans and 14 14 14 15 15 15 16 1718-18 18 18 18 19 19 19 19'1 22 22 22 22 23 23 Diablo Canyon SSER 27 vi
I.
Introduction "This supplement No.
27 to the Safety Evaluation Report addresses the changes to the Diablo Canyon, Unit 1 Operating License DPR-76 resulting from the issuance of Amendment 10 to the license which allows full power operation.
Specifical-ly, all'revisions to the license conditions and to the full power Technical Specification that resulted from the issuance of Amendment 10 are discussed below.
The affected license conditions are -identified in Section II and dis-cussed in Section IV.
The changes to the Technical Specifications are iden-tified in Section III and discussed in Secti'on V.
II.
Revised License Conditions The Facility Operating License has been revised in the following areas:
1.
Maximum power level is changed to 100 percent of full power 2.
3.
4.
5.
7.
8.
The license was revised to include referencing Supplement No; 23 on fire protection.
License condition 2. C. 8. q has been revised to stipulate a submittal date
'or the detailed control room design review summary report, a completion date for operator training on the Safety, Parameter Display System and revised emergency operating procedures and.an implementation date for revised emergency operating procedures.
A new license condition has been added to require resolution of masonry wall issues and implementation of all walls fixes prior to startup follow-ing the first refueling, License condition 2. C.(9) has been revised as a result of meetings with PG8E and ACRS on the Seismic Design Bases Revalidation Program.
A new license condition has been added to require'ommitments regarding the guidelines of Section
- 5. 1. 2 through
- 5. 1.6 of NUREG-0612 in regard to control of heavy loads prior to startup following"the first refueling outage.
Reporting of violations to the Commission has been revised to conform to 10 CFR 50.73(b), (c) and (e).
In regard to emergency preparedness, reference is made to the applicabil-ity of the provisions of 10 CFR 'Section 50.54(s)(2) should the NRC find that the lack of progress in the completion of procedures in the Federal Emergency Management Agency's rule, 44 CFR Part 350, i's an indication that a major problem exists in achieving or maintaining an adequate state of preparedness.
Diablo Canyon SSER 27
9.
10.
License condition 2.C.(10) required resolution of a jet impingement issue prior to exceeding 5X power.
This issue has been satisfactorily resolved as reported in Supplement No.
24.
License condition 2.C. (11) required the resolution of seven, issues in the areas of'iping and piping supports prior to exceeding 5X power.
These issues have been satisfactorily resolved as reported in Supplement No.
25.
Consequently,. this license condition has been met.
The license has been revised to indicate the new license expiration date.
III.
Revised Technical S ecifications The Facility Technical Specifications have been revised in the following areas:
New accident monitoring instruments have been added to Tables 3.3-10 and 4.3-7 and ACTIONS b.
and c.
have been added to Specification 3.3.3.6 for this instrumentation.
These requirements were added as recommended by NRC Generic Letter 83-37.
ACTION e. to Specification 3.3.3.6 was incorporated into the Technical Specification which is in, compliance with the wording of Standard Technical Specifications (NUREG-0452, Rev. 4) and states that provisions, of, Specification
- 3. 0.4 are not applicable.
2.
ACTION b. to Specification 3.6.4. 1 was added to address the situation where both hydrogen analyzers are -inoperable.
The existing ACTION state-ment only addressed the situation in which only one hydrogen analyzer was inoperable.
This requirement,was added as recommended by NRC Generic Letter 83-37.,
3.
4.
5.
Specification 4.8.1.1.2.b.7 required "Verifying that on a loss of the diesel generator (with offsite power not available and no Safety Injection Signal), the loads are shed from the emergency busses and that subsequent reloading of the diesel generator is in accordance with the design require-ments at least once per 18 months during plant shutdown."
This requirement has been deleted as recommended by NRC Generic Letter 83-30, and the previous Items 8 through 13 now become 7 through 12.-
Specification 3.6. 1.7 specifies, in part, that operation with the purge supply and/or exhaust isolation valves open or with the vacuum/pressure relief isolation valves open up to 50'hall be limited to less than or equal to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per 365 days.
A statement was added in the BASES which states that the 200-hour limitation shall become effective beginning at initial criticality.
The value of 365 days was changed to a calendar year to clarify the intent for a fixed time period rather than a sliding time period.
n Specification 3.6.1.7, Containment Ventilation System, required clarifica-tion due to the use of the "and/or" conjunction in the wording of the requirement.
The containment ventilation system.includes three penetra-tions:
1) a purge supply line, 2) a purge exhaust line, and 3) a common vacuum/pressure relief line.
The intent of the requirement was not clear on how many and which containment ventilation system penetrations may be open simultaneously.
The requirement was clarified to indicate that any two of the three penetrations may be used simultaneously.
Diablo Canyon SSER 27
6.
7.
A new Section 3/4.4. 11 was added to the -Technical Specifications.
The new specification provides the requirements for the reactor vessel head vents.
A new Bases section was also added.
This requirement was added as recom-mended by NRC Generic Letter,83-37.
Sections 3/4.7.5, 3/4.7.6, and 3/4.9. 12 of the Technical Specifications, which relate to testing of HEPA filters and charcoal adsorbers, were clarified in accordance with the recommendations contained in NRC Generic Letter 83-13.
A new BASES section was also added for these sections.
12.
13.
Section 3/4. l. 1.3 of the Technical Specification has been revised to allow entry into Mode 1 with a positive all-rod-out moderator temperature coefficient (MTC) in conjunction with other ACTION statements which require control rod configurations ensuring that the actual MTC is always negative during operation.
It includes an ACTION c., which states that the provisions of Specification 3.0.4 are not applicable.
ACTION b. to Specification 3.4.2.2 has been added to allow limited entry into Mode 3 for required surveillance testing of,the pressurizer Code safety valves.
Specification 4.3. l. 1 has been revised to add a Note 11 to Table 4.3-1, REACTOR TRIP SYSTEM SURVEILLANCE RE(UIREMENTS, and reference it in Item 21 of the table, "Reactor Trip Breakers".
Note,11 requires that "At least once per 18 months and following maintenance or adjustment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include verification of the independence of the Undervoltage trip and Shunt trip".
Specification 3.6.2.3 has been revised to ensure that three fan coolers will be OPERABLE for post-acci.dent containment cooling after the postulated single active failure of a two-cooler, group, as required in the containment analysis basis of the Diablo Canyon FSAR, Table 6.2-5.
t Section 3/4.7.7 of the Technical Specifications has been revised to remove the snubber table and clarify the surveillance requirements in accordance with the recommendations contained in NRC Generic Letter 84-13.
A new BASES section was also added for this sect'ion.
Section
- 6. 10.2 was also revised for reporting requirements.
Section 3/4.7. 13 of the Technical Specifications has been added to provide requirements on the condition of the breakwater.
The staff stated in Section 2.4. 1 of SSER 17 that the staff would require, and the applicant agreed to, Technical Specifications to (1) monitor the condition of the breakwater, (2) implement timely corrective action when limited damage is sustained, and (3) identify the limiting condition for operation relative to the configuration of the breakwater.
These three areas are addressed by this new section in the Technical Specifications.
A new BASES was also added for this section.
14.
Table 3.3-11 for Specification 3.3.3.8 has been revised to reflect the minimum number of instruments required OPERABLE to be consistent with the requirements in the specifications.
Diablo Canyon SSER 27
15.
16.
17.
18.
19.
20.
21.
22.
Specification 6.2.2f has been'dded to require administrative control to limit overtime for unit staff.
Specification 6.5.2.8e and f which addressed the frequency of audit-ing of emergency and security plans and the associated implementing procedures has been deleted and subsequent items renumbered.
Specifications 6.2.2, 6.2.3.4, 6.3.1, 6.5.1.2, 6.5.2.2, 6.5.2.9, 6.5.2.10, 6.8.3, Table 6.2-1 and Figures 6.2-1 and 6.2-2 have been revised to reflect administrative changes within PG8E.
Relief from Specifications 3.4.1.2 and 3.7.1.3 was granted in order to conduct a natural boron circulation and dilution test.
Currently, all four reactor coolant pum'ps cannot be de-energized for a period in excess of one hour.'ince the test requires that the:four reactor coolant pumps be de-energized for periods greater than one hour, relief was granted in order to conduct this test.
During these tests the water volume in the condensate storage tank may drop below the minimum volume required by Specification
- 3. 7. l. 3 but one alternate water source could be made avai 1-able.
Relief from this requirement was granted.
Table 4.3-1 of the Technical Specifications was revised to clarify the intent of'he survei'llance requirements for the overtemperature DT instrumentation to include the RTD bypass loop flow rate.
Section 3/4.4. 1.2 of the Technical Specifications has been revised to reflect the Westinghouse analysis'or a bank withdrawal accident in MODE 3 requiring two reactor coolant loops to be in operation.
ACTION b.
has been added to allow continued operation with one loop operating provided the Reactor Trip System breakers are open.
The Bases for this section was also changed.
Specifications 3.8.3. 1 and 3.8.3.2 were revised to correct an error in the Technical Specifications which" could have permitted the battery chargers to be inoperable for an hour or possibly eight hours than the associated battery before reactor shutdown was initiated.
Specification 5.2. lh was revised to show the correct containment free volume.
23.
Sections 6.9.1.10, 6.9.1.12, 6.9.1.13, 6.9.1.14 and 6.9.2 of the Technical Specifications were revised to,reflect reo'rganization within the NRC.
24.
N Section 6.12 of the Technical Specifications has been revised to clarify the intent of th'e requirements for high radiation areas.
IV.
Discussion of Revised License Conditions Maximum Power Level Facility Operating License DPR-76 was issued on September 22, 1981, and contained a condition limiting the maximum power level'o 5 percent of rated power.
The Licensee successfully completed its low power test pro-gram on May 23, 1984.
Diablo 'Canyon SSER 27
As a result of the discovery of seismic design deficiencies, in certain
- systems, structures, and components, and NRC concerns over quality assurance, the NRC conducted a comprehensive reevaluation of these issues.
" Numerous allegations on various issues were also received by the staff.
Supplements 16, 18, 19, 20, 21, 22, 24, 25 and 26 presented the staff's evaluation of all the above areas of concerns.
Based on the staff's eval-uation,.the Atomic Safety and Licensing Board s Initial"Decision of August 1982," and on the Commission's decision on the issuance of a full power license of July 1984 ther'e are no open items that require staff evaluation to resolve prio'r to "allowing Diablo Canyon, Unit 1 to exceed 5 percent of full power with the exception of the post accident sampling system must be operable prior to exceeding 5 percent of rated power.
On this basis, the maximum power condition in the Diablo Canyon, Unit 1 license is changed by this amendment to allow full power operation.
1q 2.
Fire Protection S stem (A
endix R to 10 CFR 50)
PG8E provided a report comparing the existing and the proposed fire pro-tection features in Diablo'ari/o, Unit 1 to the technical requirements of Sections III.G, III.J., III.L, and III.0 in Appendix R of Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) in submittal dated July 15, 1983 and supplemental submittals dated September 23 and 27, October 3, 6,
11 and 14, 1983 and May 16, 1984..
Based on the staff's evaluation as reported in Supplement No. 23, the'staff required that the following modi-fications be completed prior to exceeding 5 percent'of power.
1.
chemical laboratory and offices (fire area 4-A) 2.
G bus compartment (fire area 4-A-1) 3.'
bus compartment (fire area 4-A-2) 4.
- shower, locker, and access control (fire area 4-B) 5.
corridor outside the diesel generator room (fire area 11-D) 6.
component cooling water heat exchanger (fire area 14-E) 7.
diesel generator room (fire areas TB-l, TB-2, and TB-3) 8.
unlabeled fire doors 9.
emergency lighting (fire areas/zones CR-l, 3-R, 1, 6-A-l, 6-A-2, 6-A-3', 7-A, 5-A-4, elevation
- 104, TB-'4, TB-5, TB-6, 3-g-l, 3-g-2, TB-1, TB-2,'B-3; 3-BB, elevation 100, 5-A-1, 5-A-2, 5-A-3)
By letter dated July 6, 1984 the licensee stated that all of these modifi-cations have been completed thus removing the need to make it a condition of the license.
- However, the license has been revised to include refer-ence to Supplement No.
23.
Diablo Canyon SSER 27
In Supplement No.
8 to the Diablo Canyon Safety Evaluation Report, we stated that the carbon dioxide fire suppression system in the diesel gen-erator rooms was designed to seismic category 1 requirement in order to prevent an earthquake from causing a common-mode failure, resulting in the contamination of the diesel's combustion air.
P Subsequent to the issuance of SSER No. 8, the licensee downgraded the sys-tem t'o non-seismic Category 1.
This was done after an evaluation was con-ducted which confirmed that an inadvertent discharge of carbon dioxide (CO<) would not contaminate the diesel engine combustion air, and also would not affect diesel engine or generator cooling.
The diesel engine is cooled by water circulating thorugh the fan driven radiator.
Since the radiator is located outside the protected diesel gen-erator vault and draws in air from the plant yard, the cooling of the radiator is maintained.
In the event of an inadvertent discharge of the CO~, the CO~ will be drawn through the radiator.
However, since CO~ is an effective heat transfer medium; the cooling of the diesel engine will not be significantly affected.
In addition, cooling airwill be continually drawn into the radiator from the yard during the short period that the CO~
gas may be present.
The cooling of the generators is assured by maintaining ventilation flow through the diesel generator vaults by air ducts.
Rolling fire doors lo-cated between the diesel vault and the radiator compartment control the cooling air to the generator and are designed to remain open except in the.
event of a fire.
When a fire occurs, the doors will close with the melt-ing of a fusible link or by fire detection of a Class I thermal detection circuit.
(The thermal detectors actuate the fire door linkage through use of an electrically operated frangible link.)
The rolling fire door and the associated mechanical linkages have been seismically qualified.
The rolling fire doors will not close following a design basis earthquake and will maintain ventilati,on to the generator unless there is a fire affecting that diesel generator.
The effects on combustion air of an inadvertent discharge of CO< prio~ to diesel start and during engine operation were also, examined.
For an inad-vertent discharge prior to start of the diesel
- engine, CO< will flow from the diesel vault into the radiator compartment and dissipate into the plant yard through the missile barrier.
With a relative density of approximate-ly one and one-half times that of air, the CO< will flow near the floor.
Since the diesel combustion air intake is located 16'-6" above the floor, the air intake will be exposed to clean air.
When the engine starts, it will draw clean combustion air from above the CO< and,from the yard through the missile barrier.
The radiator fan will discharge the remain-ing CO~ from the diesel vault and radiator compartment to the plant yard away from the combustion air intake.
For an inadvertent discharge with,the diesel generator operating, the CO<
will be drawn from the diesel vault into the radiator compartment and dis-charged to the yard by the radiator fan.
Diablo Canyon SSER 27
The gas drawn into the radiator has been determined to be a mixture of approximately 77K outside air and 23K generator cooling gas.,
The genera-tor cooling gas consists of the air/CO~ mixture that is present in the vault and the outside air that is drawn into the vault, Utilizing the most conservative assumption that all of the generator cooling gas ini-tially present in the vault is CO>, the resulting mixture of outside air and cooling air will be 77' air and 23K - CO~.
This gas mixture pro-vides an Oz level in excess of the diesel engine manufactorer's minimum values for operation, which are 74K air and 26K CO~ gas.
Since the radia-tor discharges on the north end on the Turbine Building (Unit 1), there will be minimal recirculation to the radiator intake at the missile barrier.
We conclude that on the basis of this analysis, an inadvertent discharge of CO< will not significantly affect the diesel engine combustion air quality and will have no adverse affect on the diesel erigine or generator cooling.
Therefore, the licensee's redesign of the carbon dioxide system to non-seismic category I requirements is acceptable.
3.
Emer enc Res onse Ca abilities
- PG8E, by submittals dated April 18, August 2, September 9, September 19, November 8, 1983 and April 30, 1984, responded to NRC Generic letter 82-33 which superceded the completion dates contained in NUREG-0696.
PG8E has provided a Detailed Control Room Design Review (DCRDR) program plan in its August 2, 1983 submittal, as required, and stated in its April 30,
- 1984, submittal that it will provide a
(DCRDR) Summary Report to the NRC by De-cember 31, 1984.
PG8E also stated that the Emergency Operations Facility (EOF) and the Technical Support Center (TSC) are functional.
The Safety Parameters Display System (SPDS) has been declared fully operational and operators initially trained.
PG8E further stated in the April 30, 1984 submittal that operator training on the SPDS and emergency operating pro-cedures would be completed by March 28, 1985 and that PGEE will implement the emergency operating procedures by March 28, 1985.
These revised pro-cedures will have been prepared in accordance with the Westinghouse Owners Group Generic Emergency
Response
Guidelines, Revision 1.
The staff has reviewed the above submittals and find that the scheduled dates for com-pletion of the above cited activities acceptable.
The license has been conditioned to require PG8E to meet above stated completion dates.
In Supplement No.
13 to the Diablo Canyon Safety Evaluation Report we stated that prior to the issuance of a full power license we would condi-tion the full power license as follows:
1.
The applicant shall evaluate all the masonry walls in accordance with the requirements of IE Bulletin 80-11'and implement any needed fixes or modifications to meet the bulletin criteria prior to the
- full power operation.
- Also, as required by the bulletin, the appli-cant shal-1'invoke applicable action statement of the plant technical specifications, if the operability of any safety-related system is judged to be in jeopardy, because of safety considerations related to masonry walls.
Diablo Canyon SSER 27
2.
3.
The fixes or modifications implemented in (1) above, shall not preclude 'the option of implementing additional modifications if directed by future staff review of the applicant's design criteria.
Prior to startup following the first refueling, the applicant shall resolve the differences between, the staff interim criteria and the criteria used by the applicant to the satisfaction of the staff and the licensee shall submit a schedule acceptable to the staff regard-ing implementation of the required wall fixes or modifications that might result from such a resolution.
By submittals dated July 22,
- 1981, and August 17, 1983 PG8E provided its evaluation of all masonry walls in accordance with the above bulletin cri-teria and the required fixes were implemented as stated in PGEE submittal dated July 2, 1984.
This satisfied the requirement of item 1 above and thus removes the need to make it a condition of the license.
Similarly with item 2 above, PGKE stated in its letter of July 2, 1984 that the fixes made as a result of item 1 will not preclude additional modifications if required.
Therefore, the requirements of item 2 have been met and obviated the need to include it as a license condition.
Item 3 has been made a condition of the license with certain modifications.
5.
Seismic Desi n Bases Revalidation Pro ram On February 23, 1984 the NRC staff recommended a license condition for the Diablo Canyon Nuclear Power Plant because'f the Advisory Committee on Reactor Safeguards (ACRS) letter of July 14, 1978 which suggested "that the seismic design of Diablo Canyon be reevaluated in about 10 years tak-ing into account applicable new information" and as a means of assessing new information on geology and seismology of coastal California that was anticipated.
On March 27, 1984, the Commission instructed the staff to meet with PG8E and the ACRS to discuss the specific elements of license condition 2.C.(9) to reevaluate the seismic design basis for Diablo Canyon.
On April.13, 1984 the Commissioners made a condition of the license the following paragraph.
PGEE shall develop and implement a State-of-the-Art Program to revalidate the seismic design bases used for Diablo Canyon.
PG&E shall submit for NRC staff review and approval the pro-posed Program Plan and proposed schedule for.implementation by January 30, 1985., The program shall be completed and final report submitted to the NRC by July 1, 1988."
t On May 8, 1984 the staff and PG8E held a meeting regarding a program for revalidation of the Diablo Canyon, seismic design basis.
The "staff described in detai 1'the specific elements that should be included in the program.
The following were among the elements described by the staff Diablo Canyon SSER 27
a.
Eval uate post 1979.ASLB, hearing informati on b.
Reevaluate selected pre-1979 data that may,be needed c.
Reevaluate the magnitude of the controlling earthquake using a multi-method approach d.
e.
Reevaluate the ground motion at the site using analysis of empirical data, theoretical numerical modeling studies, and soil-structure interaction analyses.
Assess the significance of the results of the preceding analyses, as necessary, to assess the adequacy. of the seismic margins.
f.
Submit quarterly progress reports, participate in semi-annual meetings with the staff and make, annual presentations to the ACRS.
At the conclusion of the meeting a representative of PG8E indicated that
- overall, PGRE understands the reevaluation
- program, the elements laid out by the staff, and in general concurs in the concept of the approach for the program.
On May 24, 1984, the staff and its advisor the
Phenomena and their geological and seismological consultants had a meeting.
The staff and PG8E presented their proposed seismic reevaluation programs.
On June 14, 1984 the staff and its advisors the
The staff's proposed program was presented to the ACRS.
PG8E representatives indicated that they are in general agreement with the NRC staff proposal and will submit a program plan by January 30, 1985.
In its letter to the Commissioners (dated June 20, 1984) the ACRS stated "We believe that the elements outlined in the NRC Staff's proposal will provide a suitable basis for the seismic reevaluation.
We believe also that the NRC Staff's proposal is responsive to the July 14, 1978 ACRS letter in which the ACRS suggested that the seismic design of Diablo Canyon be reevaluated in about ten years taking into account applicable new information."
They also requested that the ACRS be given the oppor-tunity to review and comment on the PG8E program plan and schedule and that the NRC staff meet with them as appropriate to discuss the evalua-tion of the PG8E work.
In its June 20, 1984 letter, the ACRS recommended consider ation of its consultants'dvice concerning the proposed seismic reevaluation, which are summarized as follows:
1.
Analyses should include inelastic response of the plant structures under strong earthquake ground motion.
2.
Near field strong motion above a thrust fault, including the possi-bility of a strong velocity pulse, shoul"d be, considered.
Diablo Canyon SSER 27
3; All components of near field strong ground motion should be included in the analysis simultaneously.
Torsional and rocking input ground motion should not be ignored.
Three-dimensional soi 1 structure interaction should be employed to provide estimates of structural response.
5.
Advantage should be taken of existing proprietary seismic profile and well data.
6.
A critical review and evaluation should be made of the regional tectonic structure as, well as>the onshore and near offshore faults at the site in light 'of the new evidence that they may connect with an underlying thrust fault.
y It is the staff's opinion that elements (1) and (2) of revised License Condition 2.C.(9) fully address ACRS consultants'ecommendations 5 and 6.
Element (3) requires that soil structure interaction effects be considered as discussed in recommendation 4.
Elements (2) and (3) which require, as
'eeded, the assessment
'of ground motion generated by an underlying thrust fault will take into.account ACRS recommendations (2) and (3).
Element (4) provides for a seismic Probabi listic Risk Assessment (PRA) of the plant.
In a seismic PRA ground motion levels beyond the design are assumed and the.consequences investigated.
The exact nature of the deter-ministic studies to be performed will be discussed with the licensee.
If it is deemed necessary to perform an inelastic analysis as prescribed in the ACRS recommendation then such an analysis will be performed.
The ACRS's final statement is "Based on the information developed in these meetings and considering the above
- comments, we find -no reason to alter the conclusions stated in the Committee's report dated July 14, 1978 re-.
garding operation of this nuclear plant."
It is the staff's,conclusion that while the requirements of the seismic conditions are being carried out by PG&E and the staff, there is no reason to modify previous conclusions on the seismic design basis.
License Con-dition 2.C.(9)'has been revised by the staff based on the results of the above cited meetings and discussions.
6.
Overhead Heav -Load Handlin S stems As a result of Generic Task A-36, "Control. of Heavy Loads Near Spent Fuel,"
a set of guidelines was developed to assure safe handling of heavy loads over structures; systems and components important to safety.
These recommendations were documented in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."
Following the -issuance of NUREG-0612, a generic letter dated December 22, 1980 was sent to all operating plants, applicants for operating lice'nses and= hol'ders of construction permits requesting that responses be prepared to indicate the degree of compliance with the guide-lines, of NUREG-0612.
The responses were to be made in two stages.
The
'first response (Phase I, Section
- 5. 1. 1 of'NUREG-0612) was to identify the load h'andling equipment within'he scope of NUREG-0612 and to describe Diablo Canyon SSER 27 10
the associated general load handling operations such as safe load paths, procedures,- operator training, special and general-purpose of=lifting---
- devices, maintenance, testing and repair of equipment and the handling equipment specifications.
The second response (Phase II) was intended to show that'ither single-failure-proof handl,ing equipment was not needed or that single-failure-proof equipment has been provided.
This Safety Evaluation-and the attached Technical Evaluation Report (TER) constitutes the staff's evaluation of Phase I.
An evaluation of Phase II is continuing and will be provided in a future evaluation'.
As indicated above, in accordance with the generic letter dated December 22, 1980, the licensee for Diablo Canyon, Unit 1 (Pacific Gas 8
Electric Company) was requested to review their provisions for handling and control of heavy loads at the Diablo Canyon facility to determine the extent to which the guidelines of NUREG-0612 are satisfied and to commit to mutually agreeable changes and modifications that would be required in order to fully satisfy these guidelines.
The-staff" and its consultants, Idaho National Engineering Laboratory (INEL) have reviewed PGLE's submittals for Diablo Canyon Unit 1.
As a result of its review, INEL has issued the attached TER.
The'taff has reviewed the TER and concurs with its findings that the guidelines of NUREG-0612, Sec-tion 5 ~ 1. 1 have been satisfied.
Consequently, the TER is incorporated as a part of this Safety Evaluation.
We conclude that Phase I of NUREG-0612 for Diablo Canyon Unit 1 is acceptable.
The licensee currently complies with the guidelines of Section
- 5. 1. 1 of NUREG-0612 (Phase I) as accepted by the staff.
This compliance provides adequate protection in the interim against concerns associated with handling of heavy loads until full compli-ance with NUREG-0612 criteria, including Phase II at the first refueling outage.
The staff review of Phase II of NUREG-0612 for Diablo Canyon Unit 1 will be the subject of a future evaluation.
Until that review is
- complete, the following condition shall be inclu'ded in the Diablo Canyon operating license:
Prior to startup following the first refueling outage PG8E shall submit commitments necessary to implement changes and modifica-tions as required to satisfy the guidelines of Sections
- 5. 1.2 through
- 5. 1.6 of NUREG-0612 (Phase II:
9-month responses to the NRC generic letter dated December 22, 1980).
7.
Re ortin of Violations To be consistent with Code of Federal regulation, license condition 2.H has been revised to require the license to report violations to the Commission" in conformance with 10 CFR 50.73(b),-(c) and (e).
a Emer enc Pre aredness A new license condition 2.C.(13) has been added on emergency pre-
'paredness.
Reference is'ade'o the applicability to the provisions of
.'"'0 CFR 50.54(s)(2) should the NRC find that lack of progress in the completion of"procedures in the Federal Emergency Management Agency's
'ule, 44 CFR Part 350, is an indication that a major problem'exists in achieving or m'aintaining an adequate state of preparedness.
Diablo Canyon SSER 27
License condition 2.c.(10) required the licensee to perform. appropriate jet impingement analyses for certain lines inside the containment.
The licensee has submitted these analyses and they have been evaluated by the staff,and found to be acceptable as stated in Section 2 of Supplement No.
24.
Therefore, license condition 2.C.(10) has been satisfied.
10.
Pi in and Pi in Su orts On March 29, 1984 the EDO directed that a comprehensive review be initiated with respect to the large and small bore piping issues raised by Mr. Isa Yin, Region III inspector assigned to review allegations at the Diablo Canyon plant.
The Diablo Canyon Piping Peer Review Group (Group) was formed in response to this direction.
The Group consisted of nine senior staff engineers from NRR, IE and the Regions expert in piping and support design and quality assurance, and one expert piping consultant.
The Group first held discussions with Mr. Yin, following their review of his inspection report, and reviewed relevant licensee responses to sec-tions of that report.
The Group then traveled to California where a
public transcribed meeting was held with the licensee.
At this meeting the licensee presented its responses to the concerns enumerated in the subject inspection report.
H Following the public meeting, members of the Group traveled to the reactor site in order to Mr. Yin might show them physical examples that represented his areas of concern.
The initial report of the Group was published on April 12, 1984.
In that report, the Group recommended seven specific actions to be required of the licensee prior to a full power licensing decision.
Those seven recommendations were the bases for the seven licensing conditions ultimately approved by the Commission when low power operation was authorized and issued as "Order Modifying License on 4/18/84."
The Diablo Canyon Peer Review Group issued its findings in Supplement No.
25 and concluded that the seven license conditions imposed on the low power license have been satisfactorily addressed by the licensee, that the past staff conclusions on the IDVP remain valid, and that the programmatic issues raised concerning onsite engineering have been resolved.
Therefore, license condition 2.C.(11) has been satisfied.
License Ex iration Date License Condition 2. K has been revised to show an April 23, 2008 license expiration date reflecting the issuance of a full power license applied for and evaluated.
12.
Com liance with Re ulator Guide 1.97 The current license condition required PGLE to submit within thirty days of issuance of the low power license, which was issued on September 22,
- 1981,
-a proposal, including -an implementation schedule, for compliance with Regulatory Guide 1.97.
PGEE has met this license condition by sub-mittal dated October 22, 1981.
Subsequently, Generic Letter 82-33 was Diablo Canyon SSER 27 12
issued to all licensees of opei ating reactors, applicants for operating
- licenses, and holders of construction permits.
This generic letter in-cluded additional clarification regarding Regulatory Guide 1.97, Revi-sion 2 relating to requirements for emergency response capability.
These requirements have been published in Supplement 1 to NUREG-0737.
In re-sponse to Generic Letter 82-33, PG8E provided submittals-dated April 18, and September 9,
1983 which the staff has reviewed and found to be accept-able.
Based on this review, the'staff concluded that PG8E provided an explicit commitment in conformance with Regulatory Guide 1.97 with the exceptio'n of certain items which were identified by PG8E.
Based on the review of these submittals, the staff has issued an interim report request-ing additional justification for certain exceptions.
In the interim, the licensee can proceed with its implementation
- schedule, which is permitted by NRC Generic Letter 82-33 since it states that NRC review is not a pre-requisite for implementation.
13.
Residual Heat Removal RHR Low Flow Alarm In Section 6.3 of Supplement No.
13 we required the installation of a low flow alarm during the first refueling outage.
In supplement No.
22 we required that the above alarm be installed prior to exceeding 5X of power.
By letter dated June 6, 1984 the licensee stated that the low flow alarm has been installed, tested and is now functional.
Therefore, the need to require a license condition to install the alarm has been. removed.
14.
Low -Tem erature Over ressure Protection In the Diablo Canyon, Units 1 and 2,
- SER, Supplement No.
6 (Section 5.2.2, July 1977),
we discussed our concern about overpressure events occurring during low temperature operation, and measures and interim analyses which would assure safe operation of the plants.
In submittal, dated May 28, 1982, the licensee has described the low temperature overpressure miti-gating system which has been installed at both units to address our long-term requirement stated'in SER Supplement
'No.
6.
The system which uses the pressurizer power operated relief valves (PORVs) for pressure mitigation, is enabled by the operator at 323'F and activates whenever the reactor coolant system (RCS) pressure exceeds 450 psig and the RCS temperature is below 323'F.
The licensee has stated that all the components of the system meet seismic Category 1 and IEEE 279 Criteria.
In its discussion, the licensee has provided a summary of the results of analyses to identify a worst case (mass addition by spuriously started safety injection pump) and to show adequacy of the system setpoint.
In a'ddition to the administrative procedures associated both with opera-tion of the system and with reducing the potential for overpressure events
"(dis'cussed "in SER Supplement No. 6), the licensee provided governing Technical Specifications for the system, From our review as discussed
- above, we have concluded that the low-temperature overpressure protection systems provided for Diablo Canyon, Units 1 and 2, meet the requirements of RSB 5-2 and are acceptable and removes the need-to make it a condition to the license.
Diablo Canyon SSER 27 13
V.
Discussion of Revised Technical S ecifications Accident Monitorin Instr umentati on 2.
The Licensee, in response to NRC Generic Letter 83-37, submitted proposed changes to the Technical Specifications by letter dated March 30, 1984.
The proposed changes are in regard to additional operability and survei 1-lance requirements for accident monitoring instrumentation which is required in order to meet the provisions of Section 11.F.l of NUREG-0737.
We have reviewed the proposed changes and find that the proposed changes conform with the provisions of Generic Letter 83-37 and Sections ll.F. 1 of NUREG-0737 and are, therefore, acceptable.
Accordingly, Tables 3.3-10 and 4.3-7 have been expanded to include radiation monitors for the contain-ment area, plant vent and main steam line, and containment sump level moni-tors.
A description of the calibration for the containment monitors was added as a footnote to Table 4.3-7 and Actions c.
and d. were added for these monitors.
Action e.
was added to be consistent with current Standard Technical Specification wording.
All these changes are indicated in the Technical Specification by marginal lines.
Combustible Gas Control By letter dated March 30, 1984 the licensee proposed a change to the Tech-nical Specifications in response to NRC Generic Letter 83-37.
The proposed change would revise Section 3.6.4. 1 of the Technical Specification to address what action should be taken when both of the hydrogen analyzers/
monitors are inoperable.
Currently, the Technical Specifications only address what action should be taken when one hydrogen analyzer is inoper-able.
The staff concludes that the proposed change conforms to the recom-mendations contained in Generic Letter 83-37 and is, therefore, acceptable.
Accordingly, Action b. (Specification
- 3. 6. 4. 1) has been added to the Tech-nical Specifications which reads as follows:
"With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."
3.
Diesel Generator Testin The Licensee, by submittal dated March 30, 1984, responded to NRC Generic Letter 83-30 requesting an amendment to the Technical Specifications in regard to testing of diesel generators.
The amendment request would revise the present Technical Specifications by deleting Surveillance Requirement
- 4. 8. 1. 1. 2. b. 7 which requires verification that on a simulated loss of a diesel generator (with offsite power not available and no.Safety Injection Signal) the loads are, shed, from the emergency busses and that subsequent loading of the diesel generator is in accordance with design requirements.
In addition, remaining surveillance requirements are renumbered accordingly.
The proposed change is recommended by NRC Generic Letter 83-30.
The Ge-neric Letter states that the additional testing required by Specification 4.8. 1. 1.2.b.7 is not consistent with GDC 17, Regulatory Guide 1. 108, and the NRC Standard Review Plan.
The Generic Letter states that Specifica-tion 4.8. 1. 1.2:b."7 should be deleted.
Finally, the proposed change will Diablo Canyon SSER 27 14
reduce the number of ambient fast starts - a concern expressed in NRC Generic Letter 83-41.
Based on our review of the licensee's
- request, we have concluded that deletion of Specification 4.8. 1. 1.2.b.7 of the Tech-nical Specification is acceptable.
4.
Containment Ventilation S stem Specification 3.6. 1.7 specifies, in part, -that operation with the purge supply,and/or exhaust isolation valves open or with the vacuum/pressure relief isolation valves open up to 50'hall-be limited to less than or equal to 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per 365 days.
The licensee, by letter dated March 30,
- 1984, requested that Section 3.6. 1.7 be revised to add a footnote to indi-cate that the 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> limitation on operation of the above cited con-tainment purge system to be effective beginning at initial criticality.
The licensee has stated that startup testing coincident with the com-pletion effort for heatup-related modifications necessitated a large number of personnel working inside containment.
Atmospheric contaminants,
-such as carbon monoxide, from the curing of epoxy paints and debris from welding and grinding have created the need for frequent containment purg-ing to protect the worker's health and safety.
As a result, frequent con-tainment purging was being performed during the period prior to initial criticality.
Consequently if the aforementioned purging is included in the 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> limitation for the first year of operation, plant operation may be affected.
Since no fission products were present prior to initial criticality, there was no release of radioactive effluent to the environment.
Based on our evaluation of the licensee's
- proposal, we concluded that the purpose of the 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> limitation was to minimize the potential for release of radioactive effluents following an accident and that the pro-posed revision to indicate that the 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> limitation to be effective beginning at initial criticality is acceptable.
The value of 365 days was changed to a calendar year to clarify the intent for a fixed time period rather than a sliding time period.
5.
Containment Ventilation S stem Specification 3.6. 1.7, Containment Ventilation System, required clarifica-tion due to the use of the "and/or" conjunction in the wording of the LCO.
The containment ventilation system includes three penetrations:
1) a purge supply line, 2) a purge exhaust line, and 3) a common vacuum/pressure relief line.
The intent of the LCO was not clear on how many and which contain-,
ment ventilation system penetrations may be open simultaneously.
Subse-quently, this issue was clarified during telephone conversation between the Licensee and NRC in which the licensee was informed that any two out of the three penetrations may be used at the same line.
To avoid any mis-interpretation, Specification 3.6. 1.7 has been revised to include the above clarification.
6.
Reactor Vessel Head Vents PGKE, in response to NRC Generic Letter 83-37, by submittal dated Decem-ber 14, 1983 requested a revision to the Technical Specifications that Diablo Canyon SSER 27 15
7.
would add a new section which would'specify,limiting conditions for opera-
.tion and surveillance requirements for the'reactor vessel head vents.
A new BASES section would also be added.
The proposed changes are also re-quired by Item II.B. 1 of NUREG-0737.
The reactor"vessel head vent system, which has been installed, enhances safety by providing the capability to vent non-condensibles from the reactor coolant system whose presence could inhibit core cooling during natural circulation.
The revision adds a new Section,3/4.4. 11 which stipulates the limiting condition for operation and surveillance 'requirements for the reactor vessel vents.
A new BASES was
.also added.
The staff has evaluated the above revisions and find them to be consistent with the guidelines contained in Generic Letter 83-37 and Item 11.B. 1 of NUREG-0737 and are, therefore, acceptable.
Accordingly, these revisions have.-been incorporated into the Technical Specifications.
Testin of HEPA Filters and Charcoal Adsorbers In response to NRC Generic Letter 83-13, the license'in its February 29, and March 21, 1984 submittals.requested to revise Sections 4.7.5.1, 4.7.6. 1, and 4.9. 12 as they relate to the testing of HEPA filters and charcoal adsorbers.
These filters and" adsorbers are associated with the control room ventilation system, the auxiliary building safeguards air filtration system, and the fuel handling building ventilation system.
The following revisions to Specifications 4.7.5.'1, 4.7.6. 1 and 4.9. 12 were proposed:-
a.
Revise Section 4.7..5.1c (items a and b), Section 4.7.6.1b (items b and c), and Section 4.9. 12b (items b and c) to include additional information on allowable penetration and leakage for in-place testing.acceptance criteria and allowable methyl iodide penetration in the laboratory testing criteria for carbon samples.
lt b.'evise Sections 4.7.5. 1d, 4.7.6. 1c; and 4.9. 12c to include the allowable methyl iodide=penetration in the laboratory testing criteria for carbon samples.
c.
Revise Sections 4.7.5. 1f, 4.7.6. 1e and 4.9, 12e to clarify the allowable penetration and bypass leakage.in the in-place testing=-acceptance criteria for HEPA filter banks after
'complete or partial replacement.
d.
Revise Sections 4.7.5. lg, 4.7.6. lf, and 4.9. 12f to clarify and increase the allowable penetration and bypass "leakage in the -in-place testing acceptance criteria for charcoal
'dsorber banks after complete or partial replacement.
e.
Correct the number of the section.titled "AUXILIARYBUILD-ING SAFEGUARDS AIR FILTRATION SYSTEM", from 3. 4. 7. 6 to 3/4.7.6.
The licensee also stated in its February 29, 1984 submittal that the removal efficiencies assumed in the system evaluation were as follows:
Diablo Canyon SSER 27 16
1.
Control Room Ventilation System - 95K for elemental iodine, organic
- iodide, and particulates.
2.
- Auxiliary Building Safeguards Air Filtration System - 90K for elemental iodine and particulates and 70K for organic iodide.
3.
Fuel Handling Building Ventilation System - 90K for elemental iodine and 70K for organic iodine; and 90K for particulates.
The staff has evaluated the proposed changes and find that they are con-sistent with the guidance provided by Generic Letter 83-13, Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 and the removal efficiencies specified in the Final Safety Analysis Report and are, therefore, accept-able.
Accordingly, the above cited changes have been incorporated into the Technical Specifications and the affected changed identified by marginal lines.
8.
Moderator Tem eratures Coefficient MTC In a letter dated June 8, 1984, the licensee proposed,a revised Moderator Temperature Coefficient Specification
- 3. 1. 1.3 for inclusion in the full power license for Diablo Canyon Unit 1.
The revision to the proposed specification excepts the provisions of Specification 3.0.4.
Specification 3.0.4 prohibits entry into an operational
- mode, unless the conditions for the limiting condition for operation are met, without reli-ance on provisions contained in the ACTION requirements.
ACTION a. of the specification is intended to allow operation in MODES 1 and 2 if Specifica-tion 3. 1. 1.3 is not met and states operation in Modes 1 and 2 may proceed provided:
1.
Control rod withdrawal limits'are established and maintained suf-ficient to restore the MTC to less positive than 0 delta k/k/4F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
These withdrawal limits shall be in addition to the insertion limits of Specification
- 3. 1.3.6.
2.
The control rods are maintained within the withdrawal limits estab-lished above until a subsequent calculation verifies that the MTC
=
has been restored to within its limit for the all rods withdrawn condition.
3; In lieu of any other report required by Specification 6.9. 1, a
Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within '10 days describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.
Nevertheless, addition of the proposed
- change, which adds ACTION c.,
"The provisions of Specification 3.0.4 are not applicable;"
makes the intent of the Specification unambiguous.
The proposed change is therefore acceptable, Diablo Canyon SSER 27 17
Pressurizer Code Safet Valves The licensee, in a letter dated February 29, 1984, requested a revision to Specification 3.4.2.2 of the Technical Specifications that would add a
footnote "The provision of Specification 3.0.4 are not applicable" in or-der to perform maintenance on the pressurizer Code safety valves.
Cur-rently, Specification 3.0.4,would require that surveillance following maintenance on these valves be performed prior to entering MODE 3.
The staff has evaluated the requested revision and conclude that limited entry time into MODE 3 is required in order to perform the cited activities.
Accordingly, ACTION b.
has been added to Speci,fication
- 3. 4. 2. 2 to allow limited operations in MODE 3 to perform the setting of the pressurizer Code safety valves for the lift setting value.
10.
Reactor Tri Surveillance Specification 4.3. 1.1 has been revised to add a Note ll to Table 4.3-1, REACTOR TRIP SYSTEM SURVEILLANCE RE(UIREMENTS, and reference it in Item 21 of the table, "Reactor Trip Breakers."
This revision is consistent with interim surveillance required by the NRC staff for Westinghouse plants since the Salem ATWS event.
These Westinghouse plants include V.
C.
- Summer, McGuire, Callaway, and Catawba.
This generic issue is under continuing review by the staff.
Fan Coolers 12.
The licensee, by letter dated June 6, 1984, requested a revision to Spe-cification 3.6.2.3 of the Technical Specifications.
The revision would require that three fan coolers would be operable for containment cooling following a loss-of-coolant accident assuming the postulated single active failures of a two-cooler group which is required by the containment analysis basis of the Diablo Canyon Final Safety Analysis Report (FSAR),
Table 6.2.5.
The staff has evaluated the above. request and finds it to be consistent with the FSAR criteria and is acceptable.
The Technical Specification has been revised accordingly.
Snubbers 13.
By submittals dated November 4, November 10, and December 14, 1983 the licensee proposed a revision to Section 3/4.7.7 of the Technical Specifi-cations that would remove the snubber table and clarify the surveillance requirements.
The staff has reviewed the above submittals and conclude that the revisions are consistent with the guidelines contained in NRC Generic Letter 84-13 and, therefore, are acceptable.
Accordingly, Tables 3.7-3 and 3.7-4 have been deleted and the surveillance requirements have been modified.
A new BASES section was also added for this section.
Section
- 6. 10.2 was also revised for reporting requirements.
Flood Protection Breakwater In Section 2.4. 1 of Supplement No.
17 to the Staff's Safety Evaluation
- Report, we required, and the licensee agreed to, Technical Specifications Diablo Canyon SSER 27 18
e 14.
15.
to (1) monitor the condition of the breakwater,,
(2) implement timely cor-rective action when limited damage is sustained,,and (3) identify the limit-ing condition for operation relative to the configuration of the break-water.
Section 3/4.7. 13 has been added to the Technical Specifications addressing the above issues.
A new BASES has also been added.
Fire Protection By submittals dated May 10, 1982 and May 2, 1983, the licensee proposed a
revision to Table 3.3-11 of the Technical Specifications which lists the minimum number of fire detection instruments that shall be OPERABLE.
The revision is purely editorial and does not reduce the minimum number of instruments required to be OPERABLE but merely clarifies what the minimum number is.
For example, instead of stating the minimum number of instru-ments OPERABLE is 1 of 1 or 10 of 12 it is being revised to state the min-imum number of instruments OPERABLE is 1 or 10 instead of the wording above.
Limits on Workin Hours By letters dated January 11, 1983 and June 6, 1984, the licensee responded to NRC Generic Letter 82-16 and,requested that the Technical Specifica-tions be revised to include more stringent control on overtime.
The staff
'has evaluated the above request and finds it to be consistent with the guidelines provided in Generic Letter 82-'16 and is, therefore, acceptable.
Specification 6.2.2f has been added to reflect the above administrative controls on overtime.
16.
Fre uenc of Auditin Emer enc and Securit Plans and Im lementin Procedures 17.
By submittal dated December 17, 1982 the licensee requested that Specifica-tion 6. 5. 2. 8 be revised to require a more stringent surveillance program that would require auditing of the emergency and security plans and their implementing procedures once every twelve months instead of once every twenty-four months currently allowed by the Technical Specifications.
Moreover this change is required to conform with the regulations i.e.,
10 CFR 50. 54(t) and 10 CFR 73.40(d), respectively.
Since 10 CFR 50. 54(t) and 10 CFR 73.40(d) require that auditing of both the emergency plan and security plan and their implementing procedures be conducted on a twelve-month interval, the licensee by letter dated June 8, 1984 requested that reference to the frequency of auditing of the above plans in the Technical Specifications be deleted since they are required, by law.
We conclude that deletion of the reference to the frequency of auditing in the above areas in the Technical Specifications is acceptable.
Administration Controls By letters dated January ll, 1983, January ll, 1984, June 6, 1984, July 18 and July 19, 1984 the licensee requested revisions to the Technical Speci-fication in the area of administrative control.
The change would revise Figure 6. 2-1 to indicate that:
t l.
The Senior Vice President, Facilities Development is now the "Execu-tive Vice President, Facilities and Electric Resources Development,"
reporting to the Chairman of the Board and Chief Executive Officer.
Diablo Canyon SSER 27 19
2.
The Manager, guality Assurance reports to the Executive Vice Presi-dent, Facilities and Resources Development.
3.
The position of Manager, Nuclear Projects is deleted, and the Onsite Safety Review Group reports to the Technical Assistant to -the Vice President, Nuclear Power Generation.
Figure 6.2-2 would be revised to indicate that:
4 The Onsite Safety Review Group reports to the Technical Assistant to the Vice President, Nuclear Power Generation.
This group previously reported to the Manager of Nuclear Projects.
The "Power Plant Engineer" is now the "Assistant Plant Manager, Tech-nical Services," reporting to the Plant Manager.
The "equality Control Supervisor" is now the "equality Control Manager,"
reporting to the Plant Manager.
7.
The "Technical Assistant to the Plant Manager" is now the "Assistant Plant Manager, Support Services."
8.
A new position called "Personnel and General Services Manager" is added.
9.
The titles of various supervisors have been upgraded to "Manager."
Section 6.2.3.4, "ONSITE SAFETY REVIEW GROUP (OSRG),
AUTHORITY," would be revised to show that:
10.
The OSRG reports to the Technical Assistant to the Vice President, Nuclear Power Generation; it previously reported to Manager, Nuclear Projects.
Section 6.5. 1.2, "PLANT STAFF REVIEW COMMITTEE (PSRC),-COMPLETION," would be revised to indicate that:
11.
The members'itles agree with their new titles in Figure 6.2-2.
Section 6.5.2, "GENERAL OFFICE NUCLEAR POWER PLANT REVIEW AND AUDIT COM-MITTEE (GONPRAC);" would be revised to indicate that:
12.
In Section 6.5.2.2, "COMPOSITION," the title "Manager, Nuclear Pro-jects" is replaced with "Project Manager, Diablo Canyon.".
13.
In Sections 6.5.2.9, Authority," and 6.5.10, "RECORDS," the title "Senior Vice President, Facilities Development, is replaced with "Executive Vice President, Facilities and Electric Resources Develop-ment" (see also change 1).
The new title of "Executive Vice President, Facilities and Electric Resources Development,"
and the new reporting relationship with the Chairman of the Board (changes 1 and 13) are the result of a realignment of responsibility at the executive level at PG8E.
Diablo Canyon SSER 27 20
The rise in reporting level of the Manager, guality Assurance (change 2) further 'strengthens the licensee's commitment to quality assurance for Diablo Canyon by increasing the independence of the gA function.
The deletion of references to the Manager,'uclear Projects (changes 3,
4 10 and 12) was made necessary when the Manager of Nuclear Projects'osi-tion was incorporated into the overal'l PG8E/Bechtel Project Organization and no longer existed as a separate entity.
In the GONPRAC organization, the change from "Manager, Nuclear Projects:
to "Project Manager, Diablo Canyon" has no effect on operations, since the individual involved and his responsibilities are the same.
The remaining changes in Unit organization (changes 5 through 8, and 11),
are necessary because of the significant increase in plant staff over the last three years and the addition of several new functions and work groups at the plant.
At the same time, the changes simplify and streamline the reporting relationships, promoting safety and efficiency of operation.
The staff concludes that the administrative controls
- changes, which relate to organizational structure, meet the staff position i n Section
- 13. 1 of the Standard Review Plan and are acceptable.
Accordingly, Specifica-tions 6. 2. 2,
- 6. 2. 3. 4, 6. 3. 1, 6. 5. 1. 2, 6. 5. 2. 2, 6. 5. 2. 9,
- 6. 5. 2. 10,
- 6. 8. 3, Table 6. 2-1 and Figures
- 6. 2-1 and 6. 2. 2 have been revised to reflect the above changes.
18.
Natural Circulation and Boron Dilution Test By letters dated February 10 and August 16, 1983, the licensee requested one-time relief from Specifications 3.'4. 1.2, 3.4. 1.3; and 3.7. 1.3 in order to conduct a natural circulation and boron mixing test that the staff re-quired the licensee to perform.
The proposed relief would permit, during the performance of the test, all four, reactor (RCP) pumps and bo'th resid-ual heat removal (RHR) pumps to be de-energized for a period exceeding the current one-hour limit while in MODE 3 (hot standby) and MODE 4 (hot shutdown).
The proposed relief would also allow the condensate storage tank (CST) level to be below that specified in current Technical Specification
- 3. 7. 1. 3.
With respect to the relief from Specifications 3.4. 1.2 and 3.4. 1.3, the Diablo Canyon boron mixing and natural circulation tests will be conducted in accordance with staff reviewed and approved test procedures that assure the core is continually cooled and an unsafe condition does not occur.
The system temperatures and pressures will be carefully monitored and recorded throughout the test.
The RCP and RHR pumps will be available for operation throughout the test and can be immediately started at any time should the need aris'e for forced reactor coolant system flow.
With respect to the CST level, the staff notes that one of the main purposes of the test is to determine the quantity of condensate required to cool the reactor coolant system to the RHR system entry conditions.
While detailed thermal hydraulic calculations have been performed to determine this quan-tity, the test will actually measure the amount of condensate necessary.
Diablo Canyon SSER 27 21
19.
Thus, the test is both a verification of the calculations and a check of
. the system performance.
It may be necessary to allow the CST level to go below the minimum normally required to derive an accurate measurement of the condensate actually expended.
Absent the relief from this requirement, makeup to the CST would have to be continually aligned and flowing, thus adding uncertainty to the measurement of the condensate expended.
An alternate supply of makeup to the CST (and thus to the steam generators) will be available throughout the test.
Technical 'Specification
- 3. 7. l. 3 has been modified to require that during the
- test, the fire water tank and its flow path to the auxiliary feedwater pumps shall be operable prior to and throughout the test.
The staff will ensure that the test. procedures, which are under review, ensure no loss of suction to the auxiliary feedwater pumps.
for the reasons discussed
- above, the staff concludes that the proposed one-time relief from Specifications 3.4. 1.2, 3.4. 1.3, and 3.7. 1.3 is acceptable.
Overtem erature hT There was the potential for misunderstanding whether or not the RTD bypass loop flow rate should be included in the channel, calibration of the over-temperature hT instrumentation.
Note 12 was added to Table 4.3-1 of the Technical Specifications on overtemperature hT to clarify the intent of channel calibration to include the RTD bypass loop flow rate as part of channel calibration.
20.
Reactor Coolant S stem MODE 3 Section 3/4.4. 1.2 of the Technical Specifications has been changed to re-flect the safety analysis performed by Westinghouse for a bank withdrawal accident while in MODE 3 which assumes two reactor coolant loops in opera-tion.
Previously one loop operating'was assumed to provide sufficient heat removal.
The BASES for this section was also revised.
21 D.C.
Sources The staff detected an error in the Action Statement for Station Batteries in NUREG-0452, Revision 4, "Standard Technical Specifications for Westing-house Pressurized Water Reactors.-"
This error exists in the Diablo Canyon Technical Specifications and one other Westinghouse pressurized water re-actor.
The error is in the Action statement for Specifications 3.8.3. 1 and 3,8.3.2 which was inadvertently changed by splitting the batteries and chargers into two different allowable outage times.
This resulted in per-mitting the chargers to be inoperable for an additional hour, or possibly eight hours before reactor shutdown would be initiated.
The staff has therefore revised the Action Statement for Specification 3.8.3. 1 and 3.8.3.2 to correct this error.
Containment Free Volume Section 5.2. 1h of the Technical Specifications, was revised to reflect the actual design feature for the net free volume of the containment building.
Diabl o Canyon SSER 27 22
23.
NRC Or anizational Chan es As a result of organizational changes within the NRC it was necessary to revise Sections 6.9.1.10, 6.9.1.12, 6.9.1,13, 6.9.1.14 and 6.9.2 to re-flect such changes.
24.
Hi' Radiation Area There is the potential for the misinterpretation that the footnote to Specification
- 6. 12. 1 also applies to Specification
- 6. 12.2.
Such an inter-pretation was not and is not the staff's intent.
Revisions have been made to Specifications
- 6. 12, 1 and 6. 12.2 to clarify the intent of the require-ments for high radiation areas and correct minor errors.
Diablo Canyon SSER 27 23
NRC FORM 335 HIPB3)
U,S NUCLEAR REGULATORY COMMISSION BIBLIOGRAPHIC DATA SHEET I
REPORT NUMBER JANipned Oy TIOC, edd Vol. Ho,, rl enyl NUREG-.Q675 Supplement No. 27 3, TITLE ANO SUBTI'TLE Safety Evaluation Report related to the operation of Diablo Canyon Nuclear Power Plant, Units 1 and 2
Docket Nos.
50-275 and 50-323 6, AUTHORISI 2
Leave bbnk
- 4. RECIPIENT'S ACCESSION NUMBER 5, DATE REPORT COMPLETED JULY 1984
- 7. DATE REPORT ISSUED MON'TH YEAR 6, PERFORMING ORGANIZATIONNAME ANO MAILINGADDRESS (Include Zrp Codel Division of Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 11, SPONSORING ORGANIZATIONNAME ANO MA!UNG ADDRESS (Include Zsp Codd Same as 8. above 9, PROJECT/TASK/WORK UNIT NUM R
- 10. FIN NUMBER 124. TYI'E OF REPORT Technical 12b. PERIOD COVERED llncluSlve desnl 13, SUPPLEMENTARY NOTES 14 ABSTRACT l200 vvords or lessl Supplement No.
27 to the Safety Evaluation Report for Pacific Gas and Electric Company's application for licenses to operate Diablo Canyon Nuclear Power Plant, Units 1
and 2 (Dockets 50-275 and 50-323),
has been prepared by the Office of Nuclear Reactor Regulation of the U. S. Nuclear Regulatory Commission.
This supplements reports on the independent design verification program
( IDVP) for Diablo Canyon and conditions contained in Amendment No.
10 to the Operating License.
154, KEY WORDS ANO DOCUMENT ANALYSIS 15b, DESCRIPTORS 16 AVAILABILITYSTATEMENT Unlimited I 7 SECURITY CLASSIFICATION UNCLASSIFIED 19 SECURITY CLASSIFICATION UNCLASSIFIED IB NUMBER OF PAGES 20 PRICE
1
Appendix A
RiiJ R EG-0675 Suoplernent No 27 APPENDIX A CONTROL Oi HEAVY LOADS AT NUC EAR POWER PANTS PACIFIC GAS AND ELECTRIC DIABLO CANYON UNIT 1
(PHASE I)
Docke'o.
SG-275 Techni ca'I Eval uati onReport Author S.
A. Jensen Principal Technical.Investioator T.
H.
S ickley Published July 1984 EGSG Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the
" U.S.
Nuclear Regulatory Con:mission Office of Nuclear Reactor Reoulation FIN NO. A6457
A8STRACT The Nuclear Regulatory Commission (NRC) has requested that all nuclear plants e'ther operating or under construe ion submi a response of complian y with NUREG-06'2, "Control of Heavy Loads at Nuclear Power Plan.s."
EgG daho, Inc.,
has contrac.ed wi.h the NRC to ev'aluate
.he responses of.hose plants presently under construction.
This report contains EG8G's evaluation and recommendations for Oiablo Canyon Uni l.