IR 05000387/1987023
| ML17146B114 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/08/1988 |
| From: | Murphy K, Wen P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17146B113 | List: |
| References | |
| 50-387-87-23, NUDOCS 8801200260 | |
| Download: ML17146B114 (7) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-387/87-23 Docket No.
50-387 License No.
NPF-14 Licensee:
Penns lvania Power 8 Li ht Com an 2 North Ninth Street Allentown Penn s 1 vani a 18101 Facility Name:
Sus uehanna Unit
Inspection At:
Berwick Penns lvania Inspection Conducted:
December 7-11 1987 Inspector:
P.
C. Men, Reactor Engineer I/7 88 date Approved by:
Mi K. G. urp, A
in Chief Special Test Programs Section, Engineering Branch, DRS dat Ins ection Summar:
Ins ection on December 7-11 1987 Ins ection Re ort No. 50-387/87-23 Areas Ins ected:
Cycle 4 startup physics testing program and power ascension tests.
Results:
No violations were identified.
Note:
For acronyms not identified, refer to NUREG-0544,
"Handbook of Acronyms and Initialisms."
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DETAILS 1.0 Persons Contacted Penns lvania Power
& Li ht Com an PP&L J. Blakeslee, Assistant Plant Superintendent R. Boesch, Reactor Engineer
"J.
R. Doxsey, Reactor Engineer Supervisor B. Forgie, Reactor Engineer
- M. M. Golden, Acting Technical Supervisor
"J. J.
Graham, Senior Compliance Engineer D. Karchner, Reactor Engineer
"C. Mar kley, Outage Test Coordinator
- D. F.
McGann, Compliance Engineer
- H. J.
Palmer, Jr.,
Supervisor of Operations
- R. J.
Prego, gA Supervisor - Operations
"D. Roth, Compliance Engineer
- Denotes those present at the exit meeting on December 11, 1987.
The inspector also contacted other licensee staff member s in the course of the inspection.
C cle 4 Reload Safet Evaluation and Startu Test Pro ram The Cycle 4 reload contains 764 fuel assemblies, including 240 fresh 9x9 assemblies fabricated by Advanced Nuclear Fuels Corporation (ANF), 296 once burned ANF 8x8 assemblies, 192 twice burned ANF 8x8 assemblies, and 36 initial core General Electric (GE) 8x8R assemblies.
This mixed vendor reload design, safety analyses, and the required Technical Specifications (TS) changes were submitted to the NRC for review.
This reload submittal was found acceptable (letter from W.
R. Butler (NRC) to H.
W. Keiser (PP&L), dated October 9, 1987).
The inspector reviewed the plant operating procedures and noted that the operational limit changes associated with the new fuel loading have been incorporated in the procedures.
The information used for Cycle 4 startup physics testing was found to be consistent with the values derived from the safety analyses and appropriate TS sections.
The startup test program was conducted according to the Startup and Power Ascension Test Sequence which was approved by the Plant Operations Review Committee (PORC) (Meeting No.87-142, November 13, 1987).
Initial Criticality of Cycle 4 was achieved on November 21, 1987.
The Unit was synchronized to the grid on November 24, 1987.
Power ascension was con-tinued through the end of November, 1987.
During this inspection period (December 7-11, 1987), the Unit has achieved 100% power operatio The test procedures used during the startup physics testing were admin-istratively controlled under the Surveillance Test Program (AD-gA-422).
This Surveillance Test Program provided general administrative guidance on all plant surveillance tests'owever, it does not specifically address the type, test condit'ion, and test sequence of physics testing which is required to be performed to demonstrate the adequacy of the core charac-teristics as a result of new fuel cycle operations.
The need to have such a unique administrative program to provide better control of the startup testing was discussed with plant management.
At the exit meeting a
licensee representative agreed that a
new administrative test procedure would be generated to control all physics related tests during begin-ning-of-cycle plant startup.
This is an unresolved item, pending the completion of the licensee's administrative program.
(50-387/87-23-01)
3.0 C cle 4 Star tv Testin The inspector reviewed selected test programs and their results to verify the following:
Procedures were provided with the detailed stepwi se instructions, including Precautions, Limitations, and Acceptance Criteria; Technical content of the procedures was sufficient to result in satisfactory calibration and test; Test programs were implemented in accordance with test sequencing procedures; Provisions for recovering from anomalous conditions were provided; Methods and calculations were clearly specified and tests were conducted accordingly; Review, approval, and documentation of the results were in accordance with the requirements of the TS and the licensee's administrative controls.
The following tests were reviewed:
3 '
Control Rod Drive Scram Time Test The control rod drive (CRD) scram time test was performed in accordance with procedure SR-155-001, Scram Time Measurement of All Operable Control Rods, Revision 4.
The licensee utilized GETARS computer method for the first scram time test performed on November 21, 1987.
Ninety-seven out of a total of 185 control rods were scrammed from the black and white pattern at 8:o power.
However, due to computer software problems, the GETAR did not provide a direct
test result.
The licensee reactor engineer utilized a manual method (Technical Instruction TI-TA-010, CRD Timing, Revision 2) to calcu-late scram time from the stored GETAR computer file.
This manual method had been verified to be conservative prior to its use and had received PORC's approval (Meeting No.87-149).
The remaining control rods scram times were obtained by utilizing the conventional single rod scram test method on November 24 and 25, 1987.
The inspector verified that the licensee's rod scram manual calcu-lation method was properly verified and validated for its use.
The inspector also reviewed test results and noted that:
(1) the average scram times for all rods at various insertion levels were all within the TS limits, (2) the maximum scram insertion time from the fully withdrawn position to notch position 5 was 3.36 seconds which was well within the TS limit of 7 seconds, and (3) the average scram times in a two-by-two array at various insertion levels met TS limits.
No unacceptable conditions were identified.
3.2 Shutdown Mar in SDM Demonstration The SOM Demonstration test was performed in accordance with Procedure SR-100-003, Shutdown Margin Demonstration, Revision 6.
The test was performed on November 20, 1987, with a moderator temperature of 141'F.
A SOM of at least 0.85% hK/K was demonstrated with the strongest rod (30-31) fully withdrawn and the adjacent rod (34-27) withdrew to notch position 12.
This result met the TS SOM requirement of 0.38%
aK/K.
On November 21, 1987, the licensee performed the in-sequence SDM test in accordance with Procedure SR-100-008, SOM test In-Sequence Critical and Shutdown Margin Demonstration, Revision 2, and obtained a
SOM of 1.46% hK/K.
This result met the TS requirements.
No unacceptable conditions were identified.
3.3 Reactivit Anomal Check The inspector reviewed test procedure SR-100-006, Reactivity Anomaly Check, Revision 4, and test results obtained since the beginning of this cycle.
The inspector noted that all critical rod configurations were in good agreement with analytically predicted values, and well within the 1% hK/K TS limits.
No unacceptable conditions were identifie.4 Core Thermal Power and APRM Calibration The licensee's procedure RE-OTP-002, Core-Thermal Power Evaluation, Revision 1,
was reviewed for technical adequacy.
The inspector independently calculated the core thermal power using this procedure and plant data obtained at the 100% power level on December 9,
1987.
The calculated result was in reasonably good agreement with the licensee's results obtained from the process computer OD-3 and off-line computer as shown below:
Test Date Power Plateau Method Results MWT 12/9/87 (1330)
100%
- OD-3 Process Computer
- Off-line Computer
"Hand Calculation (By Inspector)
3289.97 3344.34 3343.97 The difference between the OD-3 method and inspector's result was due to input data differences.
The input data used by the inspector and off-line computer method were taken from the control room panel readings which were different from the values used in the OD-3 calculation.
The inspector also reviewed the surveillance results:
APRM scram and rod block settings, and APRM calibrations performed for the month of December, 1987.
The frequency of evaluation and final APRM readings were performed within the requirements as prescribed by the TS and plant surveillance procedure.
No unacceptable conditions were identified.
3.5 Thermal H draulic Limits The inspector reviewed procedure SR-100-001, Determination of Core Thermal Limits, Revision 4, and the test results performed on December 1-9, 1987.
The inspector verified by review of the POWERPLEX computer outputs that the thermal hydraulic limits (LHGR, MAPLHGR, and MCPR) were all within the TS limits during this period.
No unacceptable conditions were identified.
3.6 Local Power Ran e Monitor LPRM S stem Calibration The inspector reviewed test procedure RE-1TP-012, Calibration of LPRMs, Revision 3, for technical adequacy.
During the startup testing period, LPRM calibration was performed at 70% power level.
LPRM gains were also updated by using the TIP system and process
computer function OD-1 at 36%,
70%,
and 100% power levels.
All LPRM Gain Adjustment Factors were checked within the allowable tolerance.
No unacceptable conditions were identified.
3.7 Core Power Distribution The inspector reviewed test result of TIP Calibration which was taken on December 7,
1987, and noted that all TIP traces were in good agreement with the SIMULATE Code calculated results.
The inspector also independently verified that the measured average axial power shape from the POWERPLEX on-line core monitoring pro-gram agreed closely with the SIMULATE Code analytically predicted value.
No unacceptable conditions were identified.
3.8
~Summar Based on the Cycle 4 startup physics test result and plant operating data obtained during this inspection period at about 242 MTD/MTU cycle burnup, this cycle's core operation has not shown any abnormal behavior thus far.
The low radioactivity offgas rate (about 112 micro ci/sec at 100% power)
and favorable RCS radiochemistry trending data indicate that the current cycle fuel integrity appears to be sound.
4.0 Thermal-H draulic Stabilit The possibility of BMR thermal-hydraulic instability in certain operating regions, especially the high power/low flow corner of the power/flow map has been previously identified through testing and operating experience in the industry.
The licensee has addressed this stability issue for the mixed vendor core as follows:
(a)
(b)
Performed a cycle specific stability analysis.
The calculated result indicated that the maximum decay ratio at the region of two pumps minimum flow and APRM rod block intercept point (i.e.,
64% power/42%
flow) is 0.74.
This value shows that the normal operating condition is stable, and is therefore acceptable.
Implemented approved TS surveillance for detecting and suppressing power oscillations in the regions considered to be susceptible to potential instabilit e The inspector verified that these surveillance tests (SR-178-003 and SR-178-004)
had been performed during power ascension period.
The test results were acceptable.
No unacceptable conditions were identified.
5.0 Inde endent Calculations/Verifications The inspector performed independent calculations/verifications of Cycle 4 startup physics testing related activities.
These included the following:
Independent core thermal power calculation as described in Section 3.4.
Independent verification of the licensee's SDM demonstration test result.
Based on the rod notch position and test condition, the inspector verified the licensee had demonstrated the required SDM.
No unacceptable conditions were identified.
6.0 gAA/c I<<
'hrough document review and discussion with a gA supervisor, the inspector noted that during Cycle 3/4 refueling outage, (}A/gC had performed survei 1-lance of reactor engineering activities including:
new fuel inspection, core shuffling witnessing, and final core loading verification.
No open items were pending as results of these gA/gC surveillance activities.
However, the inspector did not find evidence that gA had an active sur-veillancee program which covered startup physics testing.
This subject was discussed with plant management at the exit meeting.
A licensee representative stated that this area would be strengthened in future startup testing activities.
The NRC inspector will review this in a future inspection.
7.0 Exit Interview Licensee management was informed of the purpose and scope of the inspection at the entrance interview.
The findings of the inspection were periodically discussed with licensee personnel and were summarized at the conclusion of the inspection on December ll, 1987.
Attendees at the exit interview are denoted in paragraph l.
At no time during this inspection was written material provided to the licensee.
Based on the NRC Region I review of this report and dis-cussions held with the licensee representatives at the exit, it was determined that this report does not contain information subject to
CFR 2.790 restrictions.