ML20115A920
| ML20115A920 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 07/01/1996 |
| From: | Link B WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| VPNPD-96-041, VPNPD-96-41, NUDOCS 9607090115 | |
| Download: ML20115A920 (3) | |
Text
,n-
- a 5
Wisconsin Electnc POWER COMPANY 231 W Michym. PO Bon 2046. Mawaukee. WI 53201-2046 (414)221-2345 VPNPD-96-041 July 1,1996 Document Control Desk US NUCLEAR REGULATORY COMMISSION Mail Station PI-137 Washington, DC 20555
Dear Sir / Madame:
DOCKETS 50-266 AND 50-301 REOUEST FOR EXEMPTION FROM THE REOUIREMENTS OF 10 CFR 50.60 OVERPRESSURE MITIGATING SYSTEM SETP0l'NT POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 In accordance with the provisions of 10 CFR 50.12, " Specific Exemptions," Wisconsin Electric Power Company, licensee for the Point Beach Nuclear Plant, Units 1 and 2, requests an exemption from the requirements of 10 CFR 50.60, " Acceptance Criteria for Fracture Prevention Measures for Light Water Nuclear Power Reactors for Normal Operation." We are requesting this exemption to allow the application of the ASME Section XI Code Case N-514, " Low Temperature Overpressure Protection," in determining the power operated relief valve (PORV) setpoint for low temperature overpressure protection (LTOP).
- Specifically,10 CFR 50.60 requires all power reactors to meet the criteria in Appendix G, " Fracture Toughness Requirements," to 10 CFR 50. Appendix G to 10 CFR 50 requires pressure-temperature limits for the reactor vessel to be at least as conservative as those obtained in using the methods of analysis with margins of safety as established by Appendix G of the ASME Boiler and Pressure Vessel Code Section XI.10 CFR 50.60(b) stipulates that an exemption is required to implement alternate methods to those specified in Appendix G to 10 CFR 50, 10 CFR 50.55a, " Codes and Standards," incorporates the use of the ASME Code Section XI, and includes addenda through the 1988 Addenda and editions through the 1989 Edition. Because Code Case N-514 was not published in those addenda and editions, we request an exemption for 10 CFR 50.60 to use N-514 in detennining the LTOP setpoint applicable to Point Beach. We are aware that exemptions to allow the use of Code Case N-514 were previously granted to Florida Power & Light Company for Turkey Point Nuclear Plant, Units 3 and 4,on May 11, 1993; to Duke Power Company for McGuire Nuclear Station, Units I and 2, on September 30,1994; and to Virginia Power Compan) 'ar Surry Power Station, Units 1 and 2, on October 31,1995.
As specified in 10 CFR 50.12, the Conunission may grant exemptions from the regulations when special circumstances are present. We believe the requested exemption satisfies the special circumstances criterion of 10 CFR 50.12(a)(2)(iv) in that operation in accordance with the requested exemption will result in a benefit to the public health and safety. Our evaluation supporting this detenninstion follows.
9607090115 960701
~
PDR ADOCK 05000266
/
hlh D P
PDR AsuksidayofHiemshiEnew Ovywatian
af Document Control Desk i
July 1,1996 Page 2 In 1976, we were requested by the NRC Staff to design and install a plant system to mitigate the consequences of pressure transients at low temperatures. He design of the PBNP system utilizes the installed, redundant PORVs with a low pressure setpoint that is enabled after reactor coolant pressure falls below the setpoint and prior to the reactor coolant system temperature falling below the LTOP enabled temperature. The system design is based on a reference design developed by Westinghouse. The maximum allowed setpoint of 425 psig (reference PBNP Technical Specification 15.3.15.A.I.a) was determined using the methodology developed by Westinghouse.
Our design and setpoint were accepted in a Safety Evaluation Report (SER) prepared by the NRC Staff dated May 20,1980, which accompanied Amendments 45 and 60 to DPR-24 and DPR-27, respectively. This setpoint i
was established to ensure that the ASME Section XI, Appendix G operating limits for PBNP would not be exceeded by a pressure transient at low temperatures.
In 1993, Wisconsin Electric learned that the methodology used to determine the LTOP system setpoint did not account for the differential pressure across the reactor core during reactor coolant pump operation. He pressure l
input to LTOP is sensed at two locations in the reactor coolant system. These are at the reactor coolant system hot leg and at the pressurizer. With both reactor coolant pumps operating, the pressure at the core midplane may be as j
much as 63 psig higher than at the pressure sensing points.
At that time, we reviewed the plant-specific setpoint calculations performed for PBNP and determined that these calculations contained the same error. When the differential pressure across the reactor core with two reactor coolant pumps operating was considered with the current LTOP setpoint, it was determined that the pressure at the reactor vessel midplane could exceed the ASME Section XI, Appendix G, limits when reactor coolant temperature is below approximately 152 F.
To account for the differential pressure across the reactor core, administrative requirements were implemented in 1993 to restrict operation of more than one reactor coolant pump to temperatures above 160 F. Plant operation with administrative restrictions on the operation of reactor coolant pumps places an unnecessary burden on plant operators to ensure safety limits are maintained. In fact, the restriction on reactor coolant pump operations was identified as a causal factor to a LTOP actuation at Point Beach Unit 1 on March 31,1996. Increased LTOP l
actuations may have an adverse safety impact because increased actuations also increase the possibility of a PORV failure to rescat following an actuation. There is also the potential for confusion if the pump removed from sersice i
is required to support specific plant evolutions. We believe this burden is unnecessary and can be allesiated by the application of Code Case N-514.
i i
ASME Code Case N-514 allows LTOP actuation setpoint such that the ASME Section XI, Appendix G limits are j
not exceeded by more than ten percent. Application of this Code Case at PBNP would allow operation of PBNP l
without a restriction on the number of operating reactor coolant pumps. His reduces the potential for unnecessary
)
LTOP system actuations and a possible PORV failure. Herefore, the potential for adverse safety impact is also
~
reduced, resulting in a net benefit to the public health and safety. This satisfies the special circumstances criterion j
specified in 10 CFR 50.12(a)(2)(iv).
l We believe that the requested exemption also satisfies the special circumstances entenon of 10 CFR 50.12(a)(2)(ii) l in that application of the rule is not necessary to achieve the underlying purpose of the rule. The basis for the LTOP system is to preclude reactor cociant system pressure from exceeding the ASME Section XI, Appendix G l
limits at temperatures when: the reactor vessel material toughness is reduced. As documented in the bases for Code i
I 1
ya
,.es
1 e
Documentation Control Desk July 1,1996 Page 3 Case N-514, establishing an LTOP setpoint which limits reactor coolant system pressure to no more than 110%
of the Appendix G limits provides essentially the same margin of safety as ASME Section XI, Appendix G.
Application of Code Case N-514 maintains an acceptable margin of safety while maintaining operational margins for reactor coolant pump operation at low temperatures and pressures. Setpoints established in accordance with N-514 will also minimize unnecessay actuation of protection system pressure relieving devices. Therefore, establishing an LTOP actuation setpoint using the Code Case N-514 criteria satisfies the underlying purpose of the ASME Code,10 CFR 50.60, and 10 CFR 50.55a which is to ensure nuclear power plant systems and components are designed, maintained, and tested to ensure an acceptable level of safety. This satisfies the criterion of 10 CFR 50.12(a)(2)(ii).
We request that this exemption from the requirements of 10 CFR 50.60 be processed in an expeditious manner. We are scheduled to conunence a maintenance and refueling outage of Unit I on March 29,1997. We request that this exemption be processed prior to that time in order that both reactor coolant pumps can remain operational at low temperature and pressureconditions.
Please be advised that a Technical Specification Change Request (TSCR) regarding Specification 15.3.15 is currently in preparation. The changes to be proposed by this upcoming TSCR are dependent upon NRC approval of the exemption request and apthorization to utilize ASME Code Case N-514 provided by this submittal.
If you have any questions, pleasdcontact us.
Sincerely,
%g
\\
gl}p %w l
l&
Bob Li5k Vice President Nuclear Power JF cc:
NRC Resident inspector NRC Regional Administrator, Region Ill