ML17290A856
| ML17290A856 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/29/1993 |
| From: | Clifford J Office of Nuclear Reactor Regulation |
| To: | Parrish J WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| TAC-M87076, NUDOCS 9401050081 | |
| Download: ML17290A856 (9) | |
Text
Docket No. 50-397 December 29, 1993 Hr, J.
V. Parrish (Hail Drop 1023)
Assistant Managing Director, Operations Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352
Dear Hr. Parrish:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION, POWER UPRATE REVIEW (TAC NO. H87076)
On July 9,
- 1993, you submitted a request to amend the Facility Operating License and the Technical Specifications for the Washington Public Power System Nuclear Project No. 2.
The proposed amendment would increase licensed power from 3323 HWt to 3486 HWt with extended load line limit and a change in safety relief valve setpoint tolerance.
Enclosure 1 is a list of the additional information required by our mechanical engineering branch reviewers.
Please provide the requested information promptly to enable our review on this portion of the submittal to continue on schedule.
This request for information affects fewer than 10 respondents.
Therefore, it is not subject to Office of Management and Budget review under Pub.
L.96-511.
Sincerely, Original.signed by Sheri R. Peterson for James W. Clifford, Senior Project Manager Project Directorate V
Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/enclosure:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 29, 1993 Docket No. 50-397 Hr. J.
V. Parrish (Hail Drop 1023)
Assistant Hanaging Director, Operations Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352
Dear Hr. Parrish:
', j
SUBJECT:
RE(UEST FOR ADDITIONAL INFORHATION, POWER UPRATE REVIEW (TAC NO. H87076)
On July 9,
- 1993, you submitted a request to amend the Facility Operating License and the Technical Specifications for the Washington Public Power System Nuclear Project No. 2.
The proposed amendment would increase licensed power from 3323 HWt to 3486 HWt with extended load line limit and a change in safety relief valve setpoint tolerance.
Enclosure 1 is a list of the additional information required by our mechanical engineering branch reviewers.
Please provide the requested information promptly to enable our review on this portion of the submittal to continue on schedule.
This request for information affects fewer than 10 respondents.
Therefore, it is not subject to Office of Hanagement and Budget review under Pub.
L.96-511.
Sincerely,
~~~ V~P~
Enclosure:
As stated cc w/enclosure:
See next page ames W. Clifford, Senior Project Hanager Project Directorate V
Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
C
ENCLOSURE I RE VEST FOR ADDITIONAL INFORMATION PROPOSED AMENDMENT FOR POWE UPRATE ND A CHANGE IN SRV SETPOINT TOLERANCE WASHINGTON NUCLEAR PLANT NO.
2 Mechanical Engineering Branch (Section 2.5. 1)
The evaluation did not address the effects of the power uprate such as the increase of the bottom head pressure and the high pressure scram setpoint, on the structural and functional integrity of the control rod drive system (CRDS).
Please state the basis for determining the acceptability of the CRDS regarding compliance with the design code.
The information provided should include the code edition, the code allowables, the calculated maximum stresses, deformation, and fatigue usage factor for the uprated power conditions, and assumptions used in the calculations.
(Section 2.5. I)
This section states, "The flow required for CRD cooling and driving are assured by the automatic opening of the system flow control valve, thus compensating for small increases in reactor pressure.
Prior to implementation of power uprate the flow control valves will be assessed to ensure that they are capable of operating within their acceptable range."
Provide a discussion covering the licensee's planned action if the flow control valves are not capable of operating within an acceptable range and information on similar CRD
- systems, of the WNP-2 vintage, that have been tested or operated under the uprate system flow and dome pressure.
(Sections 3.3.2) - It appears that the evaluation of reactor internals did not address the code and edition used for evaluating stresses and allowables for the reactor vessel and internals.
Please provide such information and list the maximum stresses, fatigue usage factor and location of highest stressed areas for both the current design and the uprated power conditions.
(Section 3.3.2)
The shroud was evaluated for the changes in reactor internal pressure difference (RIPD) and acoustic loads resulting from the proposed power level uprate.
The licensee states, "the shroud stresses due to the power uprate RIPD values are below their allowables; and therefore, the shroud is acceptable for the power uprate."
Provide a discussion that includes the effects of the acoustic loads on the shroud's acceptability for the power uprate.
(Section 3.3.2)
The evaluation of the jet pumps did not address the jet pump holddown beams.
Some recent holddown beam failures have been reported.
Provide information and a discussion regarding the holddown beams under the proposed power uprate operating conditions with regard to acceptability of stress levels and fatigue considerations.
(Section 3.5)
This section states that simplified elastic-plastic methods were implemented for the feedwater nozzle and that the code requirements regarding these methods were met.
Please provide a
detailed discussion on the methodology, assumptions and compliance with the code including edition, and the code allowables used.
(Section 3.5)
This section states that the piping in the primary reactor coolant pressure boundary (RCPB) systems were evaluated for compliance with ASHE code stress criteria and fatigue limits.
Interface loads to system components such as valves, RPV nozzles,
- guides, penetrations and piping suspension devices were also evaluated.
Provide a discussion regarding analysis
- methods, assumptions,and compliance with their code of record for normal, upset and faulted conditions.
The discussion should include the code and Edition used for evaluating the
- stresses, displacements and fatigue usage for the power uprate.
The evaluation should also include a discussion regarding the acceptability of pipe supports and anchorages.
(Section 3.5)
Please provide an evaluation of the increased SRV.
discharge dynamic loads and the increased HSIV closure dynamic loads on the main steam line piping.
Please also provide an evaluation of the increased fluid dynamic loads on the closure capability of the various safety related valves in the plant.
(Sections
- 3. 12)
State the code and edition used for the power uprate evaluation of balance-of-plant (BOP) piping and pipe supports including anchorages.
List the limiting BOP piping systems and components with respect to the maximum stresses and safety margin as a result of the power uprate.
(Sections 3.12.1, 10.1 and 10.1.1)
The licensee did not provide an evaluation regarding the effects of the power uprate on the design basis analyses of the high energy line break locations, and pipe-whip and jet impingement loads.
Please provide such an evaluation which includes methodology and assumptions used in the analyses and the results in comparison with the design basis allowables.
(Section 3. 12.2)
Evaluations were performed only for the supports of the systems with the highest loading increases, such as main steam, reactor core isolation cooling (RCIC), recirculation and feedwater systems.
Provide a discussion on how the results of these analyses were applied to the other systems which may have smaller safety margins and lower loading increases.
(Section
- 4. 1.2.2)
This section states that the SRV containment dynamic loads were reanalyzed for the power uprate conditions and a 3X SRV setpoint tolerance.
Please provide a discussion which compares the effects of the SRY loads on various safety related piping and mechanical equipment for the design basis and power uprate conditions.
14.
(Section 10.2) - This section only discusses the effects of power uprate on the environmental qualification of equipment, but not dynamic qualification.
For safety-related equipment, the dynamic qualification should also be addressed with respect to SRV events, annulus pressurization and jet loads in the context of power uprate.
15.
(Section 10.2.3)
This section states that the mechanical component nozzle loads and support loads are not significantly affected by the increased temperature,
- pressure, and flow conditions for power uprate.
Please provide an evaluation of the other important aspects of mechanical component design including the increased fluid induced loads on component internals.
16.
(Section 10.3)
This section briefly describes the power uprate initial testing program.
Please describe in more detail how this program will verify the original licensing requirement to perform tests in accordance with Standard Review Plan Section 3.9.2, especially regarding the increased flow induced dynamic loads.
17.
(TS Section 3/4.4.2)
The Technical Specification (T/S) tolerance for the safety/relief valves (SRVs) is being revised to +/-3X.
- However, the T/S Surveillance Requirements for the SRVs should also include a
footnote to require that the valves be reset to within +/-I/ prior to returning the valves to service.
This will provide necessary assurance that the amount of setpoint drift outside the acceptable range is minimized.
This is discussed in the licensing topical report NEDC- : 31753P, "BWROG In-Service Pressure Relief Technical Specification Revision Licensing Topical Report",
as approved by the NRC.
Reference:
"Power Uprate with Extended Load Line Limit Safety Analysis for WNP-2" NEDC 32141P, Class III, June 1993. (proprietory)
~ " Hr. J.
V. Parri sh O
WPPSS Nuclear roject No.
2 Washington Public Power Supply System (WNP-2)
CC:
Hr. J.
H. Swailes WNP-2 Plant Hanager Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352-0968 G.
E.
C. Doupe, Esq.
(Hail Drop 396)
Washington Public Power Supply System 3000 George Washington Way P.
0 ~
Box 968
- Richland, Washington 99352-0968 Hr. Warren Bishop, Chairman Energy Facility Site Evaluation Council P. 0.
Box 43172 Olympia, Washington 98504-3172 Hr. Alan G. Hosier (Hail Drop PE20)
WNP-2 Licensing Hanager Washington Public Power Supply System P. 0.
Box 968
- Richland, Washington 99352-0968 Hr. D. Coleman (Hail Drop PE20)
Acting Regulatory Programs Hanager Washington Public Power Supply System P.
0.
Box 968
- Richland, Washington 99352-0968 Regional Administrator, Region V
U.S. Nuclear Regulatory Commission 1450 Haria Lane, Suite 210 Walnut Creek, California 94596 Chairman Benton County Board of Commissioners P. 0.
Box 190
- Prosser, Washington 99350-0190 Hr. R.
C. Barr U. S. Nuclear Regulatory Commission P. 0.
Box 69
- Richland, Washington 99352-0968 H.
H. Philips, Jr.,
Esq.
Winston 8 Strawn 1400 L Street, N.W.
Washington, D.C.
20005-3502
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