ML20215E120

Safety Evaluation Supporting Util 830419 & 0830,840119, 851022 & 860930 Responses Re Conformance to Reg Guide 1.97. Plant Design Acceptable W/Exception of Neutron Flux Variable

ML20215E120
Person / Time
Site:	Yankee Rowe
Issue date:	12/09/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20215E104	List: ML20215E101 ML20215E120 ML20215E126
References
RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 8612170297
Download: ML20215E120 (6)

Text

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO CONFORMANCE TO REGULATORY GUIDE 1.97 YANKEE NUCLEAR POWEP STATION DOCKET NO.50-029 INTRODUCTION AND

SUMMARY

Yankee Atomic Electric Company, licensee for the Yankee plant, was requestec' by Generic Letter 82-33 to provide a report to the NRC describing how the post-accident monitoring instrumentation meets the guidelines of Regulatory Guide 1.97 as applied to emergency response facilities. The licensee's responses specific to Regulatory Guide 1.97 were provided on April 19, and August 30, 1983. Additional information was provided by letters dated January 19, h 1984, October 22, 1985 and September 30, 1986.

A detailed review and technical evaluation of the licensee's submittals was performed by EG8G Idaho, Inc., under contract to the NRC, with general supervision by the NRC staff. This work was reported by EG8G in the technical evaluation report (TER), "Conformance to Regulatory Guide 1.97, Yankee Rowe Generating Station," dated November 1986 (attached). The staff has reviewed this report and concurs with the conclusion that the licensee either conforms to, or is justified in deviating from, the guidance of Regulatory Guide 1.97 i

for each post-accident monitoring variable except for the neutron flux variable.

EVALUATION CRITERIA Subsequent to the. issuance of the generic letter, the NRC held regional meetings in February and March 1983 to answer licensee and applicant questions and concerns regarding the NRC policy on Regulatory Guide 1.97. At these meetings, it was noted that the NRC review would only address exceptions taken to the guidance B612170297 861209 PDR ADOCK 05000029 p PDR c2

2 of Regulatory Guide 1.97. Further, where licensees or applicants explicitly state that instrument systems conform to the provisions of the regulatory guide, it was noted that no further staff review would be necessary. Therefore, the review performed and reported by EG&G only addresses exceptions to the guidance of Regulatory Guide 1.97. This safety evaluation (SE) addresses the licensee's submittals based on the review policy described in the NRC regional meetings and the con-

, clusions of the review as reported by EG&G.

5 EVALUATION The staff has reviewed the evaluation performed by our consultant contained in the enclosed TER and concurs with its bases and findings. The licensee either conforms to, or has provided an acceptable justification for deviating from, the guidance of Regulatory Guide 1.97 for each post-accident monitoring variable except for the neutron flux variable. Regulatory Guide 1.97 recommends Category 1 for this variable. The licensee indicated that the instrumentation is not g environmentally qualified and that the Vital Instrument Bus 1 (VB1) is the only power source for the neutron flux channels. Thus, the instrumentation does not meet the environmental or single failure criteria of Regulatory Guide 1.97.

The licensee states that two additional source range channels (range of 1 to 100,000 counts per second (CPS)) are available. This instrumentation is also not environmentally qualified. These two additional source range channels are powered separately from VB1; however, they do not cover the full range recommended

i' .. .-

I by the regulatory guide'and are only redundant for the two source range channels SR1 and SR2. Having VB1 as the single power source for all neutron flux channels that cover the reconsended range does not comply with the single failure criteria.

The licensee states that the control rod position indication, RCS soluble boron i concentration samples and RCS temperature will provide additional long term in-

, formation. This alternate indication is not acceptable. The first two systems 6

are Category 3 and the third system alone is not sufficient to determine neutron flux. Therefore, they are not an acceptable substitute for the neutron flux instrumentation.

The licensee states that because a scram will occur with the loss of VB1, Cate-gory 1 instrumentation is not necessary and that after a reactor scram occurs, the neutron flux decays to less than 100,000 CPS. Thus, the licensee contends that the source range instrumentation can be used to detect an uncontrolled g

approach to criticality. Recorded flux levels will verify the initial shutdown and procedures call for emergency boration on any increases in flux level.

Borated safety injection water prevents a return to criticality for loss-of-coolant accidents, and the licensee's safety analysis shows that a return to criticality is not a concern for steamline breaks because shutdown margin is maintained.

The neutron flux instrumentation is recommended fer post-accident monitoring, after a scram has occurred, for long term monitoring and for the determinaticn

e, g

' that an accident has been and is continuously successfully mitigated. The measurement of neutron flux is a key variable for detecting an uncontrolled approach to criticality, be it an inadvertent boron dilution event or another reactivity addition event. Since key variables are classified Category 1, the licensee should commit to the installation of instrumentation for the neutron flux variable that is in accordance with Regulatory Guide 1.97.

h Section 3.1 of the TER notes that the licensee committed to complete modifications identified in their August 30, 1983 submittal by the start of Cycle 18 operation. These were completed as discussed in the 1985 Annual Report for Yankee, submitted on February 28, 1986. Therefore, this section is closed.

In Sections 3.3.3 (RCS Soluble Boron Concentration) and 3.3.27 (Accident Sampling), the TER notes that these issues are being addressed by the NRC as g

part of the review of NUREG-0737, Item II.B.3. By letter dated August 1, 1985, the NRC issued its final safety evaluation on the post-accident sampling system for Yankee which concluded that all the requirements of TMI Item II.B.3 were met. Therefore, these sections are resolved.

In Sections 3.3.4 (Core Exit Temperature), 3.3.5 (Reactor Coolant Inventory) and 3.3.6 (Degrees of Subcooling), the TER notes that these variables are being addressed as part of the NRC review of NUREG-0737, Item II.F.2.

By letter' dated October 24, 1983, the staff concluded that Yankee was in compliance with the requirements of Item II.F.2 subject to specified upgrading of the inadequate core cooling instrumentation. This upgrading was completed during the 1985 refueling outage (See Annual Report, February 28,1986).

Therefore, these sections are resolved.

g For containment effluent radioactivity, effluent radioactivity-noble gas (other buildings) and common plant vent-noble gases, the licensee has committed to review the radiation levels for equipment qualification. This review is expected to be complete by May 1, 1987, at which time Yankee will decide to replace, upgrade, or supply documentation to qualify this instrumentation. The staff finds this commitment acceptable.

CONCLUSION Based on the staff's review of the enclosed technical evaluation report, and y the licensee's submittals, the staff finds that the Yankee Nuclear Power Station design is acceptable with respect to conformance to Regulatory Guide 1.97, Revision 2, except for the neutron flux variable. It is the staff's position that the licensee shall install and make operational instrumentation that meets the requirements of Regulatory Guide 1.97, Revision 2, for this variable. This instrumentation should be operational at the first scheduled outage of sufficient duration, but no later than startup following the second refueling outage after receipt of this report.

6-The licensee is requested to inforn. the Comission, in writing, of any

, significant change in'the estimated completion schedule identified in the staff's safety evaluation and when the actions have actually been completed.

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