ML19225C603
| ML19225C603 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/07/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| IEB-79-06B, IEB-79-6B, TAC-30199, TAC-8162, NUDOCS 7908010293 | |
| Download: ML19225C603 (17) | |
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June 7, 1979
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Docket No. 50-336 Mr. W. G. Counsil, Vice President Nuclear Engineering & Operations Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101 Dear Mr. Counsi'i'.
We have completed a review of the information you have provided by your letter, dated April 24, 1979, in response to IE Bulletin 79-06B for Millstone Nuclear Power Station, Unit 2 and the additional information you provided by letters dated May 24 and 31,1979. The enclosure provides an evaluation of your responsas and discusses them with respect to their specificity, completeness, and responsiveness to the intent of the requirements of this bulletin.
We have found that you have taken the appropriate actions to meet the require-ments of Bulletin No.79-06B.
However, the staff review of the Three Mile Island Unit No. 2 accident is continuing and other corrective actions may be required at a later date. To this end we plan to meet with licensees of reactors designed by Combustion Engineering on June 12, 1979 to discuss a number of matters related to the aftermath of TMI-2.
Further action by you may be determined to be necessary at that time.
Sincerely.
.h {f s)d Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors
Enclosure:
Evaluation Report cc w/ enclosure: See next page
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790805 0 420 253
Northeast Nuclear Energy Company cc w/ enclosure (s):
Willian H. Cuddy, Esquire U. S. Nuclear Regulatory Commission Day, Berry & Howard Region I Counselors at Law Office of Inspection and Enforcement One Constitution Plaza ATTN: Mr. John T. Shedlosky Hartford, Connecticut 06103 631 Park Aver.ue King of Prussia, Pennsylvania 19406 Anthony Z. Roisman Natural Resources Defense Council cc w/ enclosure (s) and incoming 917 15th Street,ri.W.
dtd.- 4/24/79, 5/24 & 31/79 Washington, D.C.
200U5 tonnecticut Energy Agency ATTN: Assistant Director, Research Mr. Lawrence Bettencourt. First Selectman Town of Waterford and Policy Development Hall of Records - 200 Boston Post Road Department of Planning and Energy Waterford, Connecticut 06385 Policy 20 Grand Street Northeast Nuclear Energy Company Hartford, Connecticut 06106 ATTN:
Supe rin te nde nt flilIstone P1 ant Post Of fice box 128 Waterf ord, Connecticut 06385 Director, Technical Assessnent Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Regino I Of fice ATTil: EIS C00RDINATGR John 2 Kennedy Federal Building Boston, Massachusetts 02203 Waterford Public Liorary Rope Terry Road, Route 150 Uaterford, Connecticut 063%
Northeast Utilities Service Company ATTN: Mr. James R. Himmelwright Nuclear Engineering and Operations P. O. Box 270 06101 Hartford, Connecticut,
420 254
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EVALUATION OF LICEilSEE'S RESPONSES TO IE BULLETIN 79-06B NORTHEAST NUCLEAR ENERGY COMPANY, ETAL.
MILLSTONE NUCLEAR P0'JER STATION, UNIT NO. 2 DOCKET NO. 50-336 Introductico By letter cited April 14, 1979, we transmitted I&E Bulletin No.79-06B to t ortheast Nuclear Energy Company (NNECO or the licensee).
This bulletin specified actions to be taken by the licersee to avoid occurrence of an event similar to that which occurred at Three Mile Island, Unit No. 2 (TMI-2) on March 28, 1979.
By letter dated April 24,1979, NNEC0 provided their respor.ses in confermance with the requirements of the Bulletin for the Millstone Nuclear Power Station, Unit No. 2 (Millstone-2). NNEC0 supplemented this response, by letters dated May 24 and 31,1979, providing clarification and elaboration of certain af the items in response to our expressed concerns.
Our evaluation of these resoonses is given below.
Evaluation In this evaluation, the paragrach nu:rbers correspond to the bulletin action iten ; and to tne licensee's response to each action item.
1.
NNECO initially reviewed the serious consequences of the TMI-2 accident with the n'a jority of their operational personnel in 420 255
. specialized training sessions presented by the Operations Supervisor.
In addition, similar presentations were made to the operators and plant management by an NRC staff team consisting of I&E and Operator Licensing Branch (OLB) representatives on April 20 and 21, 1979 NNECO provided the same train-ing (excluding the NRC portion) to any operational personnel who missed the initial lectures prior to the Millstone-2 plant startup from the refueling outage. We find that the licensee has been responsive to tnc training reouested by the reference bulletin.
2.
NNEC0 states that operating procedures have been revised to require operator verifications of conditions which could lead to voiding.
Subsecuent communications have confirmed that the procedure revisione. are complete, including review by the Plant Operations Review Committee, and that specific values of key parameters, to be monitored by the operators to assure that thJ Reactor Cooling System (RCS) remains subcooled, are provided.
a.
NNECO states that +he parameters to be checked to determine the status of co;sible core voiding, in accordance with the revised operating procedures, are prrssurizer 420 256 prcssure and hut leg temperature to determine the amount of RCS subcooling and core delta-temperature, steam generator dolta-pressure and Reactor Coolant Pump (RCP) motor current and vibration to determine the status of RCS flow. A number of control room alarms are available to warn the operating staff of off-normal conditions that could lead to core voiding.
NNEC0 has revised emergency procedures for natural circulation operations to direct the operator to monitor the egree of subcooling using the hot leg or in-core thermo-couoles versus the saturation temperature for the existing pressurizer pressure.
Guidance is provided related to the use of stear dump / atmospheric dump operation 4- _.rfunction with auxiliary feedwater flow to establish a cere ti]w pro-ducing at least a 10 F temperature gradient across the core.
D"rection is also provided to monitor the potential for void-irg by verifying a stable or decreasing core delta-tempera-U tt re of less than 50 F.
The thermocouples in the in-core ne utron detector strings may be used for monitoring the core ir both forced and natural circulation modes. We find the 1 censee's response in regards to the recognition of ocssible void fonnation durino forced or natur31 cooling mcde of operation acceptable.
420 257
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o as;ist the operators in taking appropriate actions to b.
r arevent void format'un, NNECO states that routine and non-routine operations and the resultant procedures have been revie,4ed.
For some plant operations, procedure The changes were necessary and have been implemented.
revised procedures caution against over-feeding a steam generator during water recovery so as to prevent loss of aressurizer pressure and level control. According to sub-sequent conversations with NNEC0, the procedures to be used in the event of a Loss of Coolant Accident (LOCA), Main Steam Line Rupture or Steam Generator Tube Rupture were revised to contain RCS pressure versus temperature curves 0
indicating saturation, and 50 F subcooled con-ditions. We find that the licensee has adequately addressed the operator actions required to prevent void formation.
c.
The licensee states that the appropriate operator action required to enhance core cooling in the event core voiding occurs is tr/ restore pressurizer pressure and level and reinstate RCS cooling using the steam generators.
Level is re-established using the nomal chemical and volume control system (CVCS) charging pumps or the ECCS high pressure safety injection (HPSI) system pumps, depending on RCS integrity Core cooling, provided by RCS flow 420 258 throuah the steam generators, will be maintained by tha operation of at least one RCP per loop according to the revised emergency proceriure (see Section 6-c).
NNEC0 states that the recovery of P,CS pressure and continued core cooling will assure void collapse. We find that the licensee has adequately addressed this concern of the bulletin.
3.
In the design of Millstone-2, the automatic initiation of safety injection (SI) also results in initiation of the con-tcinmenti isolation actuation signal (CIAS).
The Millstone-2 Technical Specifications (TS) setpoint values for these actuations are RCS pressure decreasing to 1600 psia or containment pressure above 5 psig. The same setpoints are used for both SI and CIAS.
NNECO states that all containment penetrations which are not required for engineered safety features operation or core cooling, and which are not isolated by locked closed containment isolation valves, are isolated by a CIAS. TS 3/4.6.3 gives the operability and surveillance recuirements for the automatic containment isolation valves.
We find that the existing containment isolation system meets the intention of the bulletin requirenents.
4 NNEC0 does not believe that it is necessary or desirable to station an individual (with no other assigned concurrent duties and in direct and continuous communication with the control room) to promptly initiate adequate auxiliary feedwater to.the steam generators during accidents at Millstone-2.
They state that 420 u> r a,
-F-bec ause of:
(1) imediate actions required by the reactor trip procedure to verify feedwater flow status; (2) complete control of the auxiliary feedwater system from the control room panel where main feecwater flow is controlled; (3) possible interfere:1ce in the movement of control operators by an unlicensed individual; (4) fifteen minutes available before auxiliary feedwater is recuired; and (5) past experience with recovery from feedwater system problems,the requirement of this bulletin item is not justified. Although the staff ajrees with many of the points raised by NNECO there is still a concern with successful auxiliary feedwater initiation for those plants which do not have automatic start. We believe that it is prudent to have an operator available in the control room able to devote his immediate attention to the feedwater control, with no other concurrent responsibilities, during transients requiring such action.
NNEC0 has documented, in the letter dated 5/31/79, that a licensed operator who has direct responsi-bility for control and 1peration of all main and auxiliary feedwatar systems will be in the main control room at all times.
They, also, provide a backup in case the licensed operator is not available.
NNEC0 further committed to document that the operator assigned to this function will at the time of a transient requiring such action take inTrediate control of the main and auxiliary feedwater systems, with no other concurrent responsibilities, until the steam generator lev?ls return to a stable condition. We find this response to the bullatin request acceptable, f'1
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-s 5.
This bullatin item relates to the operation of the power operated relief valves (PORVS) on the pressurizer, NNECO response states that indications that plant operators a.
may utilize to determine that a PORV is open are available in the control room.
They consist of a temperature indicator on the PORV commcn discharge header and cuench tank level, We find fuch instru-temperature and pressy e indication.
mentation satisfies the concern expressed in the bulletin and appropriate direction is provided by the emergency procedures.
b.
NNEC0 states that "the emergency procedure for reactor trip has been revised to direct the operator to maintain closed the isclation valve of a stuck open PORV".
In response to our questions, NNECO explained that sequential closing and possible reopening of the PORV individual block valves may be necessary to identify the leaking PORV. However, when the leaking PORV is identified, its block valve would not be reopened.
The licensee's responses indicate that appropriate procedural control of a possible leaking FORV have been i plemented.
6.
Th s bulletin item makes specific requests of licensees to en;ure that procedures and training instructions prevent the overriding of engineered safety features during accident conditions.
420 261 As a result of a reportable occurrence and in response to a.
our November 29, 1978 letter regarding containment purging during plant operation, NNEC0's indicated that appropriate procedures were recently revised to include cautions against using eo ? m nt overrides. They state, "The cautions cnly allow override if directed by approved procedures, for equipment or personnel protection, or when equipment is not needed for the operating mode". The licensee has per-formed another review, in light of the TMI-2 Accident, and ound these procedures adequate.
The licensee places special emphasis on securing the containment spray pumps, when not needed, to prevent damage to equipment such as the RCPs.
In subsequent communications s ith NNEC0, we learned that the procedure allows these pumps
'o be secured by overriding an automatic action only if the inntainment oressure is below 10 psig.
In the Millstone-2 fesign, containment air recirculation units, reduadant to the
- pray Dumps, are available during accident conditions to iandle containment cooling requirements.
The -
see's response and the abcve example indicate that e al controls, preventing the overriding of automatic pro aCtiotis of engineered safety features have been initiated in accordance with the bulletin.
420 262
. b.
NNECO stat es that applicable emergency procedures have been revised to provide the specific instruction provided by the bulletin in regards to the continuation of HPSI pump opera-tion after automatic actuation. Although this adequately addresses the requirement of the bulletin, we are providing the following clarification of the intent of paragraph 6.6.(2).
"After 50 F of subcooling has been achieved, termination of U
HPI operation prior to 20 minutes is only permissible if it has been determined that continued operation would result in an unsafe plant condition, e.g., pressure / temperature considera-tions for the vessel integrity".
In addition, NNEC0 provided instructions regarding' charging pumps operation. They state that applicable procedures have been revised and cuatain the We find same requirements as proposed by Bulletin 79-06B.
that the licensee has adequately addressed this item for HPSI and charging pump operation.
NNECO's responses say that apg.icable emergency procedures c.
have been revised to require continued operation of at least one PCP per loop during the HPSI phase following an accident. They agree to leave the RCPs running or will restart the pumps as long as the pump is providing forced flow as indicated by control room indications. We find these statements responsive to the requirements of the f ') v3
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wJJ bulletin.
The following information is provided for clarification of the intent of bulletin paragraph 6.c.
"In the event of HPSI initiation with RCP operating, at least one RCP shall remain operating in each loop as long as the pump (s) is providing fvced flow and continued operation shall not result in an unsafe plant condition, e.g., loss of seal integrity may result in system failure of greater consequence than the benefit derived from forced flow."
d.
The NNECO response states that the applicable emergency procedures have been revised to further minimize operator dependence on pressurizer level. We find that the licensee has adequately addressed this item as presented in the bulletin.
7.
The 1 censee states that all safety related valve positions, posit aning requirements and procedural controls, which ensure that he valves remain properly pcsitioned, have been reviewed and a e adequate to ensure proper oper& tion of engineered safet features. The administrative procedures for control of maintenance on safety related equipment were revised to specifically assure correct positioning of valves which were 420 264
- 11 The worked on or were used for icolation purposes.
positions of all safety related valves, except for locked valves are visually checked monthly. The positions of locled valves are visually checked prior to each startup and after any system manipulation that require their repositioning.
We find the NNEC0 statements to be an adeq,Jate response to this item of the bulletin.
3.
NNEC0 identifies all systems designed to transfer potentially radioactive gases and liquids out of the primary containment and states that all of these systems, which are not part of the engineered safety features, are automatically isolated by a CIAS.
In addition, the containment purge valves,which are open only in the refueling and cold shutdown modes of operation,are closed upon detection of high radiation in the containment.
The licensee states that to eliminate the only potential for undesirable pumping, venting or ot:1er release, a plant design change has been completed to eliminate tr ? AUTO start feature of the containnent surrp pumo.
In the event of a steam generator tube leak, the steam generator blowdown system will process radicactive water from the steam generators to the environment or aerated liquid radwdste. A CIAS or high radiation signal from the blowdown or the steam fet air ejectors will isolate blowdown, preventing an undesired release.
420 265
. Following a postulated LOCA during the recirculation phase, potentially radioactive
-r will circulate ' rom the containment sump through the t
,..nps in the auxiliary building and then back to the reactor. tie only potential leakage would be from pump seals, valve packings and other small sources. NNEC0 states that this operation would not result in any significant release.
NNEC0 also addressed the subject of administrative controls regarding the use of the manual overrides for the Millstone-2 systems. The subject of manual overrides is part of an ongoing staff review based on responses to our generic letter of November 29, 1978.
We find that the licensee has adequately addressed the bulletin concerns regarding possible release of radioactive gases or liquids from the containment.
9.
Bulletin Item 9 relates to the safety-related system maintenance and test procedures.
a.
NNECO states that the administrative procedures have been revised to specify that prior to removal of safety related systems from service the redundant system will be 420 266
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verified operable.
We find this concern of the bulletin has been properly addressed.
b.
The licensee s.ays that the procedures for maintenance and testing of safety related systems have been reviewed and changes were made to strengthen the requirement to verify operability of safety related systems prior to taking credit for the system (s) to satisfy TS requirements. We find this to be an adequate response to the request.
c.
N'4ECO respense states that a licensed operator is required to authorize all maintenance, tests, or surveillance which affect plant systems.
Prior to releasing the centrolling docunent, the operator ensures he is aware of the effect of the activity on the system or equipment. Upon com-plet on of the item, the document is returned to the operotor for acceptance or for the purpose of returning the system to service.
The NNECO response of May 24, 1979 states that the requirements for authorizing equipment maintenance, tests, or surveillance are entrusted to individuals qualified for Shift Supervisor or Supervising Control Operator positions, f7n 9'7 s
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- and that it is a corporate objective to have such personnel qualified at the NRC Senior Reactor Operator level.
The status of all safety-related equipment and Technical Specification requirements are maintained in the Shift Supervisor's log. Each oncoming shift reviews the log to keep cognizant of the status of safety-related equipment.
We find this to be an adequate response on the operating personnel notification requirements of the bulletin.
10.
NNEC0 responds that a revision to the administrative procedure on corrnunications and outside assistance has been approved.
This revision incorporates the required notifications and establishment of communication channels requested in the bulletin.
The NMEC0 response requests more specific guidance on "Immediate notification" circumstances and notes that the bulletin statement is a general statement subject to interpre-tation. We agree that the bulletin statement is, of necessity, a general statement and was prepared in light of our knowledle of the early sequence of events at TMI-2 prior to NRC notifica-tion.
We leave it to the licensee to likewise review the TMI-2 events and, using that as guidance together with his experience in routine operations and the recognition of non-routine events, prom'algate his own interpretation of prompt NRC notification, keeping in mind. NRC's role in these matters. However, we conclude 420 268
. that should a question arise in regard to NRC notification, the licensee should plan to err on the side of providing prompt noti-fication.
NNECO has reviewed the operating modes and procedures used to 11.
deal with significant amounts af hydrogen gas that could be generated and collect in the RCS or released to the containment.
They % scribe these methods v. hat they use for degats'ng the primary coolant system (the radwaste denasifier, pressurizer steam space vent, and volume control tank gas space purge).
They also described two methods for hydrogen removal fro e containment (hydrogen recombiner and containment purge).
4 Their response indicates an understanding of this concern expressed by our bulletin. We find this response acceptable.
CONCLUSION Based on our review of the information provided by the licensee to date, we conclude that the licensee has correctly interpreted IE Bulletin No.79-06B. The actions taken demonstrate his understanding of the concerns arisina from the Three Mile Island incident in eviewing their implications on his own operations, and provide added assurance for the protection of the public health and safety during plant operation.
Dated: June 7,1979
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