ML20033B128
| ML20033B128 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/09/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20033B125 | List: |
| References | |
| TASK-1.A.1.1, TASK-TM NUDOCS 8111300416 | |
| Download: ML20033B128 (5) | |
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SAFETY EVALUATION REPORT BY THE OFFICT OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 24 TO FACILITY OPERATING LICENSE NO. OPR-34 PUBLIC SERVICE COMPANY OF COLORADO FORT ST. VRAIN NUCLEAR GENERATING STATION.
DOCKET NO. 50-267 1.0 Introduction _
Fort St. Vrain, a 330 MWe high temperature gas-cooled reactor (HTGR),
was designed by the General Atoniic Company (GAC) and is operated by the Public Service Company of Colorado (PSCo) near Platteville, Colorado.
PSCo was issued a contruction permit on September 17, 1968 and submitted the Final Safety Analysis Report as Amendment 14 to its application for a construction permit and operating license for the Fort St. Vrain Nuclear Generating Station (FSV) on November 4, 1969. A Safety Evaluation Report dated Juanuary. 20, 1972 and a first supplement which was issued on June 12,'
1973 concluded that FSV can be operated, as proposed, at power levels up to 842 MWt, full 100 percent power, without endangering the health and safety of the public.
2.0 PCRV' Pressurization
2.1 Background
By telecopied letter dated October 26, 1981, the Pu.;1ic Service Company of Colorado requested temporary relief from Technical Specification LC0 4.2.7.c - FCRV Pressurization Limiting Conditions For Operaticn. The specification states that "the PCRV shall not be cressurized to more than 100 psia unless:
...(c) the interspace between the primary and secondary penetration closures are maintained at a pressure greater than primary system pressure with purified belium gas."
The requested temporary change is to operate one of the twelve steam generator modules', namely module 3-2-3, interspace pressure at slightly
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above cold reheat steam pressure.
For the past year, Public Service Company of Colorado has been performing tests on the Fort St. Vrain core to verify the absence of the power / tempera-ture fluctuation since installation of core restraint devices. Testing has been interrupted by a helium leak between the primary and secondary closure seals of steam generator module B-2-3.
The leakage of purified helium past an "0" ring, through a crack in a seal weld and into the cold reheat steam, as shown in Figure 1 and. detail A, of one of the twelve steam generator mcdules. The reason for the requested temporary change in the Technical Specifications is to permit completion of the testing fluctuation for power / temperature up to 100 percent power.
i 8111300416 811109 PDR ADOCK 05000267
2.2 Evaluation The basis for LCO 4.2.7.c is primarily to demonstrate integrity of the primary and secondary closures to eliminate the possibility of a primary coolant leak to the environment as a result of failure of the primary closure and subsequent failure of a secondary closure.
In this particular instance, the leakage path is in the interspace between the primary and secondary closures and is directed into the cold reheat steam piping.
If primary coolant helium were to escape into the interspace, high activity in the reheat steam system would result in Plant Protective System (PPS) action resulting in loop isolation. Under this PPS action, the reheat steam system would be isolated to prevent.any further primary coolant leakage into the. secondary steam system.
In addition to the PPS monitors, the reheat system is monitored by two very sensitive radiation monitors which alarm at activity levels much lower than the PPS radiation monitor (1 ci/sec compared to %1 X 10-4 ci/sec) and which in turn can be utilized by the operator as an indication of primary closure leakage for manual operator action. Therefore, any leakage into the reheat system and thus to the environment is capable of being isolated either by manual operator action or automatically by the PPS reheat radiation moriitoring instrumentation.
Public Service Company of Colorado plans to operate the Fort St. Vrain Nuclear Generating Station with the temporary relief in Technical Specifications as follows:
1.
Operate the steam generator interspace penetration for module B-2-3 as a separate entity.
2.
Reference the interspace pressure tc cold reheat steam pressure and maintain the penetration interspace at a pressure slightly above cold reheat pressure.
2.
Monitor the reheat steam radiation monitors once per shift for in-dication of primary coolant leakage into ti.e penetration interspace and into the reheat steam system.
4 Demonstrate the ability to pressurize the interspace of module B-2-3 to reactor pressure nominally every seven (7) days but not to exceed nine (9) days.
5.
Continue to monitor th.e interspace leakage path by a pressure decay test every two (2) weeks.
2.3 Evaluation and Conclusion The licensee's proposed temporary change was evaluated against the basis i
for the LCO 4.2.7.c and our review of the change request has resulted in i
cur finding that the proposed temporary change will not impose a signifi-cant adverse impact either on the health and safety of the public or on the environment. Thus, we find the licensee's proposed temporary Technical Specification change acceptable, provided the plans for operation outlined in section 2.2 are followed.
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y 3.0 Orga'nization and Administrative" Controls.
3.1 Backcround 7
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Under the provisions of 10 CFR Part 50.36(d)(3) and in order to provide reasonable assurance that all facility operations are maintained within
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the limits datermined acceptable following the implementation of the TMI-l Lessons Learned Category "A" items, the NRC prepared model Technical Specifications (TSs). These model specifications are intended to provide guidance in the scope and types of required specifications for each facility in the areas of equipment and administM tive requirements including actions the NRC considers appropriate if a limiting condition for operation cannot be met.
Our review of the Fort St. Vrain. facility resulted in the determination ~
that only the Shift Technical Advisor section should be included in the Technical Specifications. The bases for this determination are contained
'in our March 20, 1980 letter to Public Service.
Public Service Company of Colorado was nctified.of our request for Technical Specifications governing the Shift Technical Advisor by letter dated October 2, 1981.
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3.2 Evaluation'"
As requested by NRC letter dated October 2, 1981 the Public Service Company' of Colorado submitted a change to their Technical Specifications incor-porating the requirement for Shift Technical Acvisors.
The.Jechnical Specifications define the responsibility of Technical Advisors and che,ir chain of command in the organi7ation.
The Technical Specifications define the respense of the Shift Technical Advisors during working hours and "on call" after, normal working hours ~
and the respense time as one hour after an e ergency call.
The Technical Specifications define the training requirements for the Technica'l Advisor position.
The licerisee's Technical Specifications for the Shift Technical Advisor r
were reviewed against NUREG-0578, NUREG-0737 and INP0 xec'ommendations; we determined that most of the items were in close agreement with regu-lations.
Detailed evaluation is continuing at ORNL regarding bette'r: definition..
of situations requiring calling in the STA, and makings con'tingency plans to cover.the porsibility of the on-call STA being delayed. When the ORNL work is compiergJ, the staff will notify'the licensee of any required changes.
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4.0 Conclusions After reviewing the requests for Technical Specification. changes in licensee letters dated October 26, 1981 and October 30, 1981 the staff has concluded that the changes will not result in any significant environmental impact nor adversely affect t.he health and saf.ety of the.public, Therefore, we find the requests for' Technical Specification changds accsptabl4.- ~ ~ ~
Environmental Consideration Wehavedeterminedt$attheamendmen'tdo'esnotauthorizeach'angein
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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 551.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this. arrendment.
Concl'us'i on We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amencment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in comoliance 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.
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