ML19210A440

Tech Spec Change Request 36,Amend 1,supporting Licensee Request to Change App a of License DPR-50 Re Elimination of Vent Valve Flow Penalty.Certificate of Svc Encl

ML19210A440
Person / Time
Site:	Three Mile Island
Issue date:	01/31/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19210A437	List: ML19210A436 ML19210A440
References
NUDOCS 7910290660
Download: ML19210A440 (9)

Text

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UNITED STATFJi OF AMERICA NUCLEAR REGULATORY Cola 4ISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EI'ISON CCITANY This is to certify that a copy of Technical Specification Change Request No. 36, Amendment 1 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commiscion and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Penncylvania by deposit in the United States mail, addressed ac follows:

Mr. '4eldon B. Arehart Mr. Harry B. Reese, Jr.

Board of Supervicora of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletowri, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY

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PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request Nc. 36 Amendment 1 Thir Technical Specification Change Request is submitted in support of request ',o change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISON COMPANY 0

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Vice President W

Sworn and subscribed to me this day of , 1977 V /

NotaryPut[c 45 1469

Metropolitan Edison Company (Met-Ed)

Three Mile Island Nuclear Station Unit 1 (TMI-1)

Docket No. 50-289 Operating License No. DPR-50 Technical Scecification Chance Request No. 36, Amendment 1 The Licensee requests that the attached pages replace the corresponding existing Technical Specifications pages. This amendment implements the recommendation made in our letter of October 19,1976. (GQL 1407)

Reasons for Proposed Change It has been discovered that:

The fuel densification penalty was not properly incorporated into technical specifications prepared for cycle 2. Proper incorporation of this penalty would affect DNB based pressure-temperature limit curves such that they would be more restrictive.

A new rod bow penalty has been imposed as a result of Westinghouse experiments revealing a heretofore not considered phenomenon.

Babcock & Wilcox calculations confirmed that elimination of the internal vent valve bypass. flow penalty, as authorized by Nuclear Regulatory Commission letter of March 10, 1976, would compensate for this error and the imposed rod bov penalty. A more detailed discussion of =argins available for off-setting rod bow effects is presented in GQL lh07 Thus , eliminaticn of the internal vent valve flow penalty will allow continued use of present pressure temperature limits (Figure 2.3-1)

As a prerequisite for eliminating the vent valve flow penalty, the Commission required in its letter of March 10, 1976, "... testing to be conducted each refueling outage to confirm that no vent valve is stuck in an open position and that each va:.ve continues to exhibit co=plete freedom of movement." This surveillance requirement was performed during the last refueling outage, and is planned for the upcoming refueling outage. This proposed change incorporates this surveillance requirement into technical specifications.

Note: The proposed technical specification b.16 included in Technical Specification Change Request No. 13 (still under review) is no lenger needed, due to equipment modification. Therefore, Technical Specification Change Pequest No.13 has been retracted.

Safety Analysis Justfyinc P'coposed Chance Elimination of the vent valve flow penalty has been authorized by the Conmission.

Revised densification analysis indicates that the correct penalties are 5.935 DNBR (versus 1.885 in the Reload Report) and 3.h7% power peeking relative to DNBR (versus 1.065 quoted in the Relcad Repert).

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The variable low pressure trip setpoint for cycle 2 operation is based on the four pump open vent valve pressure-temperature limit curve presented in figure 2.1.3 of the present Technical Specifications (Curve 1).

Curves 2 and 3 represent the corresponding limits for 3 and 2 pump operation, respectively. Each curve is based on the assumption that the reactor is operating at the maximum achievable power level for that pump operating condition. In the original cycle 2 submittal (and in the cycle one technical specifications), Curve 1 incorporated the open vent valve penalty, while curves 2 and 3 did not. That is, the four pump limit curve was based upon operation with one vent valve open while the three and two pump limit curves assumed all vent valves remained shut. In revising Figure 2.1-3 to incorporate the corrected DUBR densification penalty, the basis for the four pump limit j curve was changed to eliminate the vent valve penalty. The combined effect l vas to move curve 1 to the right. However, in order to compensate for rod bov l effects, curve 1 was left unchanged for this submittal. The revised curves 2 and 3 have incorporated the new red bow and the increased densification penalty, therefore, the curves moved slightly to the left.

! The flux / flow trip setpoint for cycle 2 (1.08) is based on the one pump coastdown analysis. When the revised densification penalty is incorporated and the vent valve penalty is eliminated, the thermal-hydraulic limiting flux / flow setpoint is greater than 1.12 (this limit must be at least 1.11 to justify the tech spec setpoint of 1.08). It can also be shown that a thermal-hydraulic limit of 1.11 on the flux / flow setpoint can be justified by taking credit for 1/2 of the vent valve penalty.

The error found in the TMI-1, Cycle 2 DNBR densification penalty calculations resulted from the use of inconsistent heat flum (flux shape) and enthalpy rise in evaluating the DNER densification penalty. This error only affects the PT envelope and flux / flow ratio.

Based upon the above, it is determined that this change does not constitute a threat to the health and safety of the public, nor does it involve an unreviewed safety question.

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TABLE OF CONTEiTS

( Section Page 4.6 DIERGENCY POWER SYSTE4 PERIODIC TESTS h-46 h.7 REACTOR CONTROL ROD SYSTE4 TESTS h- 48 4.7.1 CONTROL ROD DRIVE SYSTDI FUNCTIONAL TESTS h-48 h.7.2 CONTROL ROD PROGRM4 VERIFICATION h-50 h.8 MAIN STEAM ISOLATION VALVES h-51 h.9 DIERGENCY FEEDWATER PU'4PS PERIODIC TESTING k-52 h.9 1 TEST h-52 h.9.2 ACCEPTANCE CRITERIA h-52 h.10 REACTIVITY ANOMALIES h-53 h.ll SITE ENVIRONMENTAL RADIOACTIVITY SURVEI k-5h h.12 CONTROL ROOM FILTERING SYSTDi h-55 4.12.1 OPERATING TESTS h-55 h.12.2 FILTER TESTS h-55 h.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE h-56 h.lh REACTOR BUILDING PURGE EXHAUST SYSTDi h-57 h.15 MAIN STEAM SYSTDI INSERVICE INSPECTION k-58 h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE h-59 5 DESIGN FEATURES 5-1 51 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2

! ( 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 l 5.3 REACTOR 5-h 5.3.1 REACTOR CORE 5-k 5.3.2 REACTOR COOLANT SYSTD4 5-4 5.h NEW AND SPENT FUEL S0TRAGE FACILITIES 5-6 5.h.1 NEW FUEL STORAGE 5-6 5.h.2 SPEiT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTE4S 5-8 1

6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-2 6.2.1 0FFSITE 6-2 6.2.2 FACILITY STAFF 6-2 6.3 STATION STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW & AUDIT 6-3 6.5.1 PLAliT OPERATIONS REVIEW COMMITTEE (PORC) 6-3 6.5.2.A MET-ED CORPORATE TECHIIICAL SUPPORT STAFF 6-5 6.5 2.B GENERAL OFFICE REVIEW BOARD (GCRB) 6-7 6.6 REPORTABLE OCCURRENCE ACTION 6-10 6.7 OCCURRENCES INVOLVING A SAFETY LIMIT VIOLATION 6-10a 6.8 PROCEDURES 6-11

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The power level trip set point produced by the power-to-flow ratio provides both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level

( trip set point produced by the power to flow ratio provides overpower D!G protection fer all modes of pump operation. For every flow rate there is a maximum pernissible power level, and for every power level there is a minimum permissible lov flow rate. Typical power level and lov flow rate combinations for the pump situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.6 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating if power is 80.7 percent and reactor flow rate is Th.7 percent or flow rate is 69 2 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 52.9 percent and reactor flow rate is h9.2 percent or flow rate is h5.h percent and the power level is E9 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation frcm the average value of the RC flow si6nal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No penalty in reactor coolant flow through the core was taken for an cpen i core vent valve because of the core vent valve survef.llance program during f

\ each refueling outage.

For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either pcVer peaking kW/ft limits or DIGR limits. The reactor power imbalance (pover in the top half of core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The pover-to-flow ratio reduces the power level trip and associated reactor power / reactor power-imbalance boundaries by 1.08 i percent for a one percent flow reduction.

b. Pump monitors The redundant pump monitors prevent the minitram core DIGR frcs

- decreasing belov 1.3 by tripping the reactor due to the loss of

- reactor coolant pu=p(s). The pu=p monitors also restrict the I power level for the number of pumps in operation. .

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' c. Reactor coolant system pressure l

l During a startup accident fro = low power or a slow rod withdrawal i frcm high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting I limit shown in Figure 2.3-1 for high reactor coolant system pressure (2355 psig) has been established to maintain the system l

pressure below the safety limit (2750 psig) for any design transient.

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h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE

( Aeolicability Applies to Reactor Internals Vent Valves.

Objective To verify that no reactor internals vent valve is stuck in t. open position and that each valve continues to exhibit freedom of move =ent.

Specification h.16.1 At intervals not exceeding the refueling interval, each reactor internals vent valve vill be tested to verify

, that no valve is stuck in the open position and that each j valve continues to exhibit freedom of movement.

Bases i t i Verifying vent valve freedom of movement insures that coolant flow does not bypass the core through reactor internals vent valves during operation and therefore insures the conservatism of Core Protection Safety limits as delineated in figures 2.1-1 and 2.1-3, and the flux / flow trip setpoint.

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l 1600 620 640 660 560 580 600 Reactor Outlet Temperature. *F REACTOR COOLANT FLOW POWER PUMPS OPERATING (TYPE OF LIMIT)

CURVE (L6S/HR) 139.8 x 106 (100",)* 112% Four Pumps (DNBR Limit) 1 66.7% inree Pumps (DN8R Limit) 2 104.5 x 106 (74.7",)

59.1% One Pump in Eacn loop (Quality Limit) 3 68.8 x 106 (49.2%)

• 106.5% of Cycle 1 Design Flow TMl-1, UNIT I, CYCLE 2 CORE PROTECTI0tl SAFETY Figure 2.1 3 1469 352