ML14175A798
| ML14175A798 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 03/28/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML14175A797 | List: |
| References | |
| NUDOCS 8404200044 | |
| Download: ML14175A798 (5) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.7 8 TO FACILITY OPERATING LICENSE NO. DPR-23 CAROLINA POWER AND LIGHT COMPANY H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 Introduction By letter dated October 14, 1984, Carolina Power and Light Company (the licensee) proposed revisions to the Technical Specifications of the H. B. Robinson Steam Electric Plant Unit'No. 2. The proposed revisions would clarify the qualification requirements for alternate members of the Plant Nuclear Safety Committee (PNSC);
make the list of technical disciplines for the Corporate Nuclear Safety Section (CNSS) Independent Review consistent with ANSI N18.7-1976; incorporate mechanical snubbers installed during the last refueling outage; incorporate Standard Technical Specifications (STS) terminology regarding the Administrative Control of High Radiation Areas; revised the Acceptance Criteria for the Integrated Leak Rate Test (ILRT); revise staffing specifications to be in accordance with requirements; and correct typographical errors and inconsistencies. The proposed revisions are discussed individually below.
ALTERNATES FOR PLANT NUCLEAR SAFETY COMMITTEE (Specification 6.5.1.6.3)
Discussion and Evaluation In their letter of October 14, 1983, CP&L proposed changes to the Administrative Controls Technical Specification 6.5.1.6.3, noting that the qualification requirements specified for alternate plant nuclear safety committee (PNSC) members did not represent all of the functional areas which compose the PNSC.
The qualification requirements specified that alternates shall, as a minimum, meet the Qualifications specified for Professional-technical oersonnel in Section 4.4 of ANSI-N18.1-1971 The licensee's proposed change which stated:
All alternates shall, as a minimum, meet equivalent qualification criteria as specified for professional-technical personnel in Section 4.4 of ANSI-N18.1-1971", was not sufficiently clear to determine how alternates not listed in Section 4.4 should be qualified. Therefore, clarifications were discussed with and agreed to by the licensee. The clarified wording is:
All alternates shall, as a minimum, meet qualification criteria specified in Section 4.4 of ANSI-N18.1-1971 for professional-technical-personnel or, for those discipl.iqes not listed in Section 4.4, the equivalent of the Section 4.4, requirement.
This clarifies the qualification requirements for alternate PNSC members and will not result in a change to facility operations. This change is administrative and therefore does not involve a significant hazards consideration.
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PDR ADOCK 05000261 P
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Conclusion With the clarifications agreed to by the licensee, the submitted administrative change to the specification is accepted and amends the existing specification.
CORPORATE NUCLEAR SAFETY SECTION INOEPENOENT REVIEW CRITERIA (Specification 6.5.2.3)
Discussion and Evaluation In their letter of October 14, 1983, CP&L proposed a change to the Administrative Controls Technical Specification 6.5.2.3 which would add the technical area of Nondestructive Testing to the list of areas in which members of the Corporate Nuclear Safety Section are required to collectively possess the experience and competence necessary to perform reviews.
The proposed change is admin'strative "in-nature and would allow the licensee to conform to the guidance of ANSI N18..7-1976 with respect to the inclusion of Nondestructive Testing in the list.
The change constitutes an additional requirement for the independent review group as listed in the Technical Specifications and does not involve a significant hazards consideration.
Conclusion The proposed change is accepted as submitted and amends the corresponding Specification.
SNUBBERS (Specifications 3.13 and 4.13)
Discussion and Evaluation In their letter of October 14,
- 1983, CP&L proposed changes to Technical Specifications which were necessitated by the addition of two safety-related mechanical snubbers to the auxiliary feedwater system.
These are the first safety-related mechanical snubbers to be installed at Robinson.
The chances to the Technical Specifications identify surveillance requirements for the 3
added safety-related shock suppressors.
The proposed changes were based on guidance provided by the NRC and do not involve a significant safety hazards consideration.
Conclusion The proposed changes are accepted as submitted and are incorporated into the Technical Specifications.
HIGH RADIATION AREA KEY CONTROL (Specification 6.13)
Discussion and Evaluation In their letter of October 14,
- 1983, CP&L proposed an administrative change to Technical Specification 6.13 which would allow for control of High Radiation Area Keys by the Radiation Control Foreman.
This proposed change conforms to the guidance used in Standard Technical Specifications by allowing these keys to be administratively controlled by the Shift Foreman on duty and/or the Radiation Control Foreman.
The change revises the administrative controls providing 'consistency with Standard Technical Specifications, enhancing control of access to High Radiation
- Areas, and allowing a reduction in the administrative burden on the Shift Foreman.
The proposed change does not involve a significant safety hazards consideration.
Conclusion The proposed change is accepted as submitted and is incorporated into the Technical Specifications.
ACCEPTANCE CRITERIA FOR INTEGRATED LEAK RATE TEST (Specification 4.4.1.1.f)
Discussion and Evaluation In their letter of October 14, 1983. CP&L proposed changes to Technical Specification 4.4.1.1.f which would increase the maximum allowable leak rate for the containment vessel during testing from 0.08 percent per day to 0.1 percent per day.
The 0.08 value represents the leakage criteria at a containment internal environmental temperature of 120 0 F, the expected air temperature during reactor operation. The 0.1 value is the equivalent leakage rate at a containment internal environmental temperature. of 263 0 F, the expected temperature of the steam-air mixture at the peak accident pressure.
The 0.08 value was used in accordance with the previously existing requirements of the AEC Technical Safety Guide (Revised Draft -
December 15, 1966) in order to correct test temperature during the Integrated Leak Rate Test to accident temperature.
- However, issuance of 10 CFR 50, Appendix J superseded the requirements of the AEC Technical Safety Guide. Appendix J does not require the 20 percent reduction in 4
leak rate from 0.1 to 0.08, but rather requires the measured leakage to be less than 75 percent of the maximum allowable leak rate. This 25 percent reduction is already included in the Robinson Technical Specifications.
Therefore, to eliminate the redundant reduction for the maximum allowable leak rate for the containment vessel, and to comply with the requirement of NRC Standard Review Plan 6.2.6 which specifies a minimum acceptable design containment leakage rate of not less than 0.1 percent per day, the licensee has proposed changing the Technical Specification 4.4.1.1.f leakage value from 0.08 percent per day to 0.1 percent per day.
The proposed change does not constitute an unreviewed safety question, nor does it involve a significant increase in the probability or consequences of an accident previously evaluated, or create the possibility of a new or different kind of accident from any previously evaluated, or involve a significant reduction in the margin of safety.
This change does not involve a significant hazards consideration.
Conclusion The proposed change is accepted as submitted and is incorporated into the Technical Specifications.
METHYL IODIDE (Specifications 3.8.2.b and 4.15.1.d)
Discussion and Evaluation CP&L proposed, in their letter of October 14,
- 1983, administrative changes to Technical Specifications 3.8.2.b and 4.15.1.d.
In each case the Specifications reference a laboratory test for "methyl iodine". The correct term for the type of laboratory testing actually required and performed is "methyl iodide". These proposed changes correct a typographical error and are purely administrative in nature.
They do not involve a significant hazards consideration.
Conclusion The proposed changes are accepted as submitted and are incorporated into the Technical Specifications.
SHIFT STAFFING (Section 6.2.2)
Discussion and Evaluation In their letter of October 14,
- 1983, CP&L proposed changes to Technical Specification 6.2.2 regarding the composition and manning of the shift staff.
The licensee's proposed changes would relax the required availability of the Shift Technical Advisor (STA),
and add requirements for an additional shift member and an additional Senior Reactor Operator during hot operations.
5 The licensee currently requires that an STA be available for duty at all times.
This requirement is more restrictive than NRC regulations and staff guidance.
NUREG-0737, Clarification of TMI Action Plan, Item I.A.1.1, Shift Technical Advisor, requires that an STA be available for duty when the plant is operating in Modes 1-4.
The proposed change in STA staffing would make the Technical Specification consistent with NUREG-0737.
The proposed added shift manning requirements will make the Technical Specification consistent with 10 CFR 50.54(m)(2) and Section I.A.1.3 of NUREG-0737.
These proposed additions constitute additional restrictions on the shift complement not presently in Technical Specifications and conform to recent revisions in the regulation as stated. These changes do not involve a significant hazard consideration.
Summary The proposed changes are accepted as submitted and are incorporated into the Technical Specifications.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4),
that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with -the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: March 28, 1984 Principal Contributor:
- 0. Price -
Region II